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{{#Wiki_filter:UNIT 1 EFFECTIVE PAGE LIST REMOVE                         INSERT 3.2/4.2-14                     3.2/4.2-14 3.2/4.2-15                    3.2/4.2-15 3.2/4.2-23                    3.2/4.2-23*
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09035b 901030 PDR ADOCK 05000>59 PNU
09035b 901030 PDR ADOCK 05000>59 PNU


TABLE 3.2.8 INSTRUHENTATION THAT INITATES OR CONTROLS THE CORE AND CONTAINHENT COOLING SYSTEHS Hinimum No.
TABLE 3.2.8 INSTRUHENTATION THAT INITATES OR CONTROLS THE CORE AND CONTAINHENT COOLING SYSTEHS Hinimum No.
Operable Per
Operable Per
~Tri )~s~l             Functi n               Tri level S t in       ~Ac icn             R  mark Instrument Channel            > 470" above vessel zero               1. Below  trip setting initiates Reactor Low Water Level                                                    HPCI.
~Tri )~s~l 2(16) 1(16)
Instrument Channel            > 470" above vessel zero.               1. Hultiplier relays initiate Reactor Low Water Level                                                    RCIC.
Functi n
Instrument Channel            > 378" above vessel zero.               1. Below trip setting initiates Reactor Low Water Level                                                    CSS.
Instrument Channel Reactor Low Water Level Instrument Channel Reactor Low Water Level Instrument Channel Reactor Low Water Level (LIS-3-58A-D, SW ¹1)
(LIS-3-58A-D, SW ¹1)
Instrument Channel-Reactor Low Mater Level (LIS-3-58A-D, SW ¹2)
Instrument Channel Reactor Low Water Level Permissive (LIS-3-184 8
: 185, SW ¹1)
Instrument Channel Reactor Low Water Level (LITS-3-52 and 62, SW ¹1)
Tri level S t in
~Ac icn
> 470" above vessel zero
> 470" above vessel zero.
> 378" above vessel zero.
> 378" above vessel zero.
> 544" above vessel zero.
A
> 312 5/16" above vessel zero.
A (2/3 core height)
R mark 1.
Below trip setting initiates HPCI.
1.
Hultiplier relays initiate RCIC.
1.
Below trip setting initiates CSS.
Hultiplier relays initiate LPCI.
Hultiplier relays initiate LPCI.
: 2. Hultiplier relay   from CSS initiates accident signal (15).
2.
2(16)  Instrument Channel-            >  378" above vessel zero.              1. Below trip settings, in Reactor Low Mater Level                                                    conjunction with drywell (LIS-3-58A-D, SW ¹2)                                                        high pressure, low water level permissive, 120 sec.
Hultiplier relay from CSS initiates accident signal (15).
1.
Below trip settings, in conjunction with drywell high pressure, low water level permissive, 120 sec.
delay timer and CSS or RHR pump running, initiates ADS.
delay timer and CSS or RHR pump running, initiates ADS.
1(16) Instrument Channel            >  544" above vessel zero.      A        l. Below trip setting permissive Reactor Low Water Level                                                    for ini ti ating si gnal s on ADS.
l.
Permissive (LIS-3-184 8 185,  SW ¹1)
Below trip setting permissive for initiating si gnal s on ADS.
Instrument Channel            >  312 5/16" above vessel zero. A        l. Below trip setting prevents Reactor Low Water Level        (2/3 core height)                          inadvertent operation of (LITS-3-52 and 62, SW ¹1)                                                  containment spray during accident condition.
l.
Below trip setting prevents inadvertent operation of containment spray during accident condition.


4   4 Sr, >~ 4 TABLE  3.2.8 (Continued)
4 4
Hinimum No.
Sr,
>~
4 Hinimum No.
Operable Per
Operable Per
~Tri S   i           Function        Tri  L  v 1    in      Action        Remark 2(18)       Instrument Channel-   1<  p<2.5 psig                    l. Below  trip setting  prevents Drywell High Pressure                                       inadvertent operation of (PS-64-58 E-H)                                             containment spray during accident conditions.
~Tri S
2(18)       Instrument Channel-   < 2.5 psig                       l. Above   trip setting in con-Drywell High Pressure                                      junction with low reactor (PS-64-58 A-D, SW &#xb9;2)                                      pressure initiates CSS.
i 2(18) 2(18) 2(18) 2(16)(18)
Function Instrument Channel-Drywell High Pressure (PS-64-58 E-H)
Instrument Channel-Drywell High Pressure (PS-64-58 A-D, SW &#xb9;2)
Instrument Channel-Drywell High Pressure (PS-64-58A-D, SW &#xb9;1)-
Instrument Channel-Drywell High Pressure (PS-64-57A-0)
TABLE 3.2.8 (Continued)
Tri L v 1
in 1< p<2.5 psig
< 2.5 psig
< 2.5 psig
< 2.5 psig Action Remark l.
Below trip setting prevents inadvertent operation of containment spray during accident conditions.
l.
Above trip setting in con-junction with low reactor pressure initiates CSS.
Hultiplier relays initiate HPCI.
Hultiplier relays initiate HPCI.
: 2. Hultiplier relay from CSS ini ti ates ace i dent s i gnal . (15) 2(18)        Instrument Channel-    <  2.5 psig                      l. Above   trip setting in Drywell High Pressure                                      conjunction with low (PS-64-58A-D,  SW &#xb9;1)-                                      reactor pressure initiates LPCI.
2.
2(16)(18)    Instrument Channel-    <  2.5 psig                      l. Above   trip setting, in Drywell High Pressure                                      conjunction with 1 ow reac to r (PS-64-57A-0)                                              water level, drywell high pressure, 120 sec. delay timer and CSS or RHR pump running, initiates ADS.
Hultiplier relay from CSS initiates ace i dent s ignal. (15) l.
Above trip setting in conjunction with low reactor pressure initiates LPCI.
l.
Above trip setting, in conjunction with 1 ow reac tor water level, drywell high
: pressure, 120 sec.
delay timer and CSS or RHR pump running, initiates ADS.


NOTES FOR TABLE       2 B
NOTES FOR TABLE 2
: 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.
B 1.
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.
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:
Action:
A. Repair in 24 hours.       If the   function is not   OPERABLE in 24 hours, take action B.
A.
B. Declare the system or component inoperable.
Repair in 24 hours. If the function is not OPERABLE in 24 hours, take action B.
C. Immediately take action       B until   power is verified on the trip system.
B.
D. No action required; indicators are considered redundant.
Declare the system or component inoperable.
: 2. In only   one trip system.
C.
: 3. Not considered     in a trip system.
Immediately take action B until power is verified on the trip system.
: 4. Requires one channel from each physical location (there are 4 locations) in the steam line space.
D.
: 5. With diesel power, each       RHRS pump     is scheduled to start immediately and each CSS   pump is sequenced     to start about     7 sec. later.
No action required; indicators are considered redundant.
: 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.,
2.
at which time the     full complement       of CSS and RHRS pumps would be operating.
In only one trip system.
: 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 line break.
3.
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.
Not considered in a trip system.
: 8. Note 1 does not apply to         this item.
4.
: 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.
Requires one channel from each physical location (there are 4 locations) in the steam line space.
BFN                                             3.2/4.2-23 Unit 1
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 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 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


NOTES FOR   ABLE     2 B     ont'd)
NOTES FOR ABLE 2
: 10. Only one   trip system   for each cooler fan.
B ont'd) 10.
: 11. In only   two of the four 4160-V shutdown boards.       See note 13.
Only one trip system for each cooler fan.
: 12. In only one of the four     4160-V shutdown boards. See note 13.
11.
: 13. An emergency 4160-V shutdown board       is considered a trip system.
In only two of the four 4160-V shutdown boards.
: 14. RHRSW pump   would be inoperable. Refer to Section 4.5.C         for the requirements of a RHRSW pump being inoperable.
See note 13.
: 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.
12.
: 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.
In only one of the four 4160-V shutdown boards.
: 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 hours, an orderly power reduction   shall be initiated and reactor power shall be less than 30 percent within four hours.
See note 13.
: 18. Not required to be     OPERABLE   in the COLD SHUTDOWN CONDITION.
13.
BFN                                       3.2/4.2-24 Unit 1
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 hours, 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


4     CORE AND CONTA     ME     COOLI G               SYSTEMS LIMITING CONDITIONS     FOR OPERATION                             SURVEILLANCE REQUIREMENTS 3.5.B   Res dug     Heat Removal S ste                             4.5.B   es dual Heat   Remova   S ste
4 CORE AND CONTA ME COOLI G SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.B Res dug Heat Removal S ste
        ~RfLRRS  . (LPCI and Containment                                  ~RHRS   iLPCP and Containment Cooling)                                                          Cooling)
~RfLRRS
: 8. If Specifications     3.5.B.1                               8. No  additional surveillance through 3.5.B.7 are not met,                                     required.
. (LPCI and Containment Cooling) 4.5.B es dual Heat Remova S ste
an orderly shutdown shall be initiated and the reactor shall be placed   in the COLD SHUTDOWN CONDITION within 24 hours.
~RHRS iLPCP and Containment Cooling)
: 9. When  the reactor vessel                                    9. When the reactor vessel pressure is atmospheric and                                     pressure is atmospheric, irradiated fuel is in the                                       the RHR pumps and valves reactor vessel, at least one                                     that are required to be RHR loop with two pumps or two                                 OPERABLE  shall  be loops with one pump per loop                                     demonstrated to be OPERABLE shall be OPERABLE. The                                      per Specification 1.0.MM.
: 8. If Specifications 3.5.B.1 through 3.5.B.7 are not met, an orderly shutdown shall be initiated and the reactor shall be placed in the COLD SHUTDOWN CONDITION within 24 hours.
diesel generators pumps'ssociated must also be OPERABLE.
8.
: 10. If the conditions of                                       10. No  additional surveillance Specification 3.5.A.5 are met,                                   required.
No additional surveillance required.
LPCI and containment     cooling are not required.
9.
ll. When there is irradiated fuel                             11. The RHR pumps on the adjacent units which supply in the reactor and the reactor is not in the COLD SHUTDOWN                                     cross-connect capability CONDITION, 2 RHR pumps and                                       shall be demonstrated to be associated heat exchangers and                                   OPERABLE per Specification valves on an adjacent unit                                       1.0.MM when the cross-must be OPERABLE and capable                                     connect  capability of supplying cross-connect                                       is required.
When the reactor vessel pressure is atmospheric and irradiated fuel is in the reactor vessel, at least one RHR loop with two pumps or two loops with one pump per loop shall be OPERABLE.
capability except as specified in Specification 3.5.B.12 below. (Note: Because cross-connect capability is not a short-term requirement, a component is not considered inoperable   if cross-connect capability can be restored to service within   5 hours.)
The pumps'ssociated diesel generators must also be OPERABLE.
BFN                                           3.5/4.5-7 Unit  1
9.
When the reactor vessel pressure is atmospheric, the RHR pumps and valves that are required to be OPERABLE shall be demonstrated to be OPERABLE per Specification 1.0.MM.
: 10. If the conditions of Specification 3.5.A.5 are met, LPCI and containment cooling are not required.
10.
No additional surveillance required.
ll.
When there is irradiated fuel in the reactor and the reactor is not in the COLD SHUTDOWN CONDITION, 2 RHR pumps and associated heat exchangers and valves on an adjacent unit must be OPERABLE and capable of supplying cross-connect capability except as specified in Specification 3.5.B.12 below.
(Note:
Because cross-connect capability is not a short-term requirement, a
component is not considered inoperable if cross-connect capability can be restored to service within 5 hours.)
11.
The RHR pumps on the adjacent units which supply cross-connect capability shall be demonstrated to be OPERABLE per Specification 1.0.MM when the cross-connect capability is required.
BFN Unit 1 3.5/4.5-7


4     CORE A D CO             COO I G SYSTEMS LIMITING CONDITIONS     FOR OPERATION               SURVEILLANCE REQUIREMENTS 3.5.B   Res dua     eat Remova S stem               4.5.B  Res dua    eat Remova  S  ste
4 CORE A D CO COO I G SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.B Res dua eat Remova S stem
        ~RHRS    (LPCI and Containment                    ~RHRS   (LPCI and Containment Cooling)                                          Cooling)
~RHRS (LPCI and Containment Cooling)
: 12. If one RHR pump   or associated             12. No  additional surveillance required.
: 12. If one RHR pump or associated heat exchanger located on the unit cross-connection in the adjacent unit is inoperable for any reason (including valve inoperability, pipe break, etc.),
heat exchanger located on the unit cross-connection in the adjacent unit is inoperable for any reason (including valve inoperability, pipe break, etc.), the reactor may remain in operation for a period not to exceed 30 days provided the remaining RHR pump and associated diesel generator are OPERABLE.
the reactor may remain in operation for a period not to exceed 30 days provided the remaining RHR pump and associated diesel generator are OPERABLE.
: 13. If RHR cross-connection   flow or         13. No  additional surveillance heat removal capability is lost,                   required.
4.5.B Res dua eat Remova S ste
the unit may remain in operation for a period not to exceed 10 days unless such capability is restored.
~RHRS (LPCI and Containment Cooling) 12.
: 14. All recirculation    pump                    14. All recirculation   pump discharge valves shall                            discharge valves shall be OPERABLE PRIOR   TO                           be tested for OPERABILITY STARTUP   (or closed if                           during any period of permitted elsewhere                                COLD SHUTDOWN CONDITION in these specifications).                          exceeding 48 hours,   if OPERABILITY tests have not been performed during the preceding 31 days.
No additional surveillance required.
3.5/4.5-8 AMENDMEN [f0. g G 9 BFN Unit  1
: 13. If RHR cross-connection flow or heat removal capability is lost, the unit may remain in operation for a period not to exceed 10 days unless such capability is restored.
13.
No additional surveillance required.
14.
All recirculation pump discharge valves shall be OPERABLE PRIOR TO STARTUP (or closed if permitted elsewhere in these specifications).
14.
All recirculation pump discharge valves shall be tested for OPERABILITY during any period of COLD SHUTDOWN CONDITION exceeding 48 hours, if OPERABILITY tests have not been performed during the preceding 31 days.
BFN Unit 1 3.5/4.5-8 AMENDMEN[f0. g G 9


4     CORE LIMITING CONDITIONS CO           COOLI FOR OPERATION G SYSTE S           t SURVEILLANCE REQUIREMENTS 3.5.C   RHR   Service Water and E er enc,         4.5.C   RHR   Service Water and Emer enc E ui  ment Coolin Wate S stems                  E  ui ment Coolin Water S stems EECWS     Continued                              EECWS    Cont nued
4 CORE CO t
: 4. One of the Dl   or D2 RHRSW                 4. No additional surveillance pumps assigned   to the RHR                       is required.
COOLI G SYSTE S
heat exchanger supplying the standby coolant supply connection may be inoperable for a period not to exceed 30 days provided the OPERABLE pump is aligned to supply the RHR heat exchanger header and the associated diesel generator and essential control valves are OPERABLE.
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.C RHR Service Water and E er enc, E ui ment Coolin Wate S stems EECWS Continued 4.5.C RHR Service Water and Emer enc E ui ment Coolin Water S stems EECWS Cont nued 4.
: 5. The standby   coolant supply capability may be inoperable for a period not to exceed 10 days.
One of the Dl or D2 RHRSW pumps assigned to the RHR heat exchanger supplying the standby coolant supply connection may be inoperable for a period not to exceed 30 days provided the OPERABLE pump is aligned to supply the RHR heat exchanger header and the associated diesel generator and essential control valves are OPERABLE.
: 6. If Specifications   3.5.C.2 through 3.5.C.5 are not met, an orderly shutdown shall be initiated and the unit placed in the COLD SHUTDOWN CONDITION   within 24 hours.
4.
: 7. There shall be at least 2 RHRSW   pumps, associated with the selected RHR pumps, aligned for RHR heat exchanger service for each reactor vessel containing irradiated fuel.
No additional surveillance is required.
BFN                                           3.5/4.5-12               AMEHOMENT g0. ygg Unit  1
5.
The standby coolant supply capability may be inoperable for a period not to exceed 10 days.
: 6. If Specifications 3.5.C.2 through 3.5.C.5 are not
: met, an orderly shutdown shall be initiated and the unit placed in the COLD SHUTDOWN CONDITION within 24 hours.
7.
There shall be at least 2
RHRSW pumps, associated with the selected RHR pumps, aligned for RHR heat exchanger service for each reactor vessel containing irradiated fuel.
BFN Unit 1 3.5/4.5-12 AMEHOMENTg0. y g g


    /4   5   CORE AND CONTAINME       COOLING SYSTEMS LIMITING CONDITIONS       FOR OPERATION                 SURVEILLANCE REQUIREMENTS 3.5.D         ui ment Area Coolers                     4.5.D   E ui ment   ea Coolers
/4 5
: 1. The equipment area    cooler                  l. Each equipment area cooler associated with each RHR                             is operated in conjunction pump and the equipment                               with the equipment served area cooler associated                               by that particular cooler; with each set of core                                 therefore, the equipment spray pumps (A and C                                 area coolers are tested at or B and D) must be                                   the same frequency as the OPERABLE at all times                                 pumps which they serve.
CORE AND CONTAINME COOLING SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.D ui ment Area Coolers 4.5.D E ui ment ea Coolers 1.
when the pump or pumps served by that specific cooler is considered to
The equipment area cooler associated with each RHR pump and the equipment area cooler associated with each set of core spray pumps (A and C
                ,be OPERABLE.
or B and D) must be OPERABLE at all times when the pump or pumps served by that specific cooler is considered to
: 2. When an equipment area cooler is not OPERABLE,     .
,be OPERABLE.
the pump(s) served by that cooler must be considered inoperable for Technical Specification purposes.
l.
E. Hi h Pressure     Coolant In ection               E. H   h Pressure   Coolant S  stem    HPC S                                      In ection   S stem   HPCIS
Each equipment area cooler is operated in conjunction with the equipment served by that particular cooler; therefore, the equipment area coolers are tested at the same frequency as the pumps which they serve.
: 1. The HPCI system     shall be                   1. HPCI Subsystem  testing OPERABLE   whenever there is                         shall be performed as irradiated fuel in the                               follows:
2.
reactor vessel and the reactor vessel pressure                         a. Simulated      Once/18 is greater than 150 psig,                             Automatic      months except in the   COLD SHUTDOWN                       Actuation CONDITION or as specified in                         Test 3.5.E.2. OPERABILITY shall be determined within 12 hours                     b. Pump            Per after reactor   steam pressure                       OPERA-          Specification reaches 150 psig from a COLD                         BILITY          1.0.MM CONDITION, or alternatively PRIOR TO STARTUP by using an                   c. Motor Oper-     Per auxiliary  steam supply.                            ated Valve     Specification OPERABILITY    1.0.MM
When an equipment area cooler is not OPERABLE, the pump(s) served by that cooler must be considered inoperable for Technical Specification purposes.
: d. Flow Rate at   Once/3 normal         months reactor vessel operating pressure BFN                                             3.5/4.5-13 Unit  1
E.
Hi h Pressure Coolant In ection S stem HPC S
E.
H h Pressure Coolant In ection S stem HPCIS 1.
The HPCI system shall be OPERABLE whenever there is irradiated fuel in the reactor vessel and the reactor vessel pressure is greater than 150 psig, except in the COLD SHUTDOWN CONDITION or as specified in 3.5.E.2.
OPERABILITY shall be determined within 12 hours after reactor steam pressure reaches 150 psig from a COLD CONDITION, or alternatively PRIOR TO STARTUP by using an auxiliary steam supply.
a.
Simulated Automatic Actuation Test Once/18 months b.
Pump OPERA-BILITY Per Specification 1.0.MM c.
Motor Oper-ated Valve OPERABILITY Per Specification 1.0.MM 1.
HPCI Subsystem testing shall be performed as follows:
d.
Flow Rate at normal reactor vessel operating pressure Once/3 months BFN Unit 1 3.5/4.5-13


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.5 4. CORE A    CO  AINMENT COOLING SYSTE    S LIMITING CONDITIONS    FOR OPERATION            SURVEILLANCE REQUIREMENTS 3.5.E  Hi h Pressure    Coo  ant In ection      4.5.E  Hi h Pressure Coolant In ection 4.5.E.1 (Cont'd)
~
: e. Flow Rate at    Once/18 150  psig      months The HPCI pump shall deliver at least 5000 gpm during each  flow rate test.
: f. Verify that        Once/Month each valve (manual, power-operated, or automatic) in the injection flow-path that is not locked, sealed, or otherwise secured in position, is in its correct* position.
: 2. If the  HPCI system  is                    2. No  additional surveillances inoperable, the reactor may                      are required.
remain in operation for a period not to exceed 7 days, provided the ADS, CSS, RHRS (LPCI), and RCICS are OPERABLE.
: 3. If Specifications    3.5.E.l
* Except that    an automatic or 3.5.E.2 are not met,                              valve capable of an orderly shutdown shall                            automatic return to its be initiated    and the                              ECCS  position when an reactor vessel pressure                              ECCS  signal is present shall be reduced to 150                              may be in a position for psig or less within 24                                another mode of hours.                                                operation.
F. Reactor Core    Isolation Coolin              F. Reactor Core Isolation Coolin
: 1. The RCICS  shall  be OPERABLE                1. RCIC Subsystem  testing shall whenever there    is irradiated                      be performed as  follows:
fuel in the reactor vessel and the  reactor vessel                            a. Simulated Auto- Once/18 pressure is above 150 psig,                              matic Actuation months except in the COLD SHUTDOWN                              Test CONDITION or as specified in 3.5.F.2. OPERABILITY  shall BFN                                      3.5/4.5-14 Unit  1


3  4     CORE AND CO       NME    COOLI  G SYS EMS LIMITING CONDITIONS     FOR OPERATION                 SURVEILLANCE REQUIREMENTS 3.5.F    Reacto  Co e    so  ation  Coo in          4.5.F  Reactor Core Iso at    o    Coo  i 3.5.F.l (Cont'd)                                  4.5.F.1   (Cont'd) be determined    within  12 hours                b. Pump                Per after reactor    steam pressure                        OPERABILITY        Specifi-reaches 150 psig from a      COLD                                          cation CONDITION or alternatively                                                  1.0.MM PRIOR TO STARTUP by using an auxiliary  steam supply.                          c. Motor-Operated      Per Valve              Specifi-OPERABILITY        cation 1.0.MM
.5 4.
: d. Flow Rate at      Once/3 normal reactor    months vessel operating pressure
CORE A CO AINMENT COOLING SYSTE S
: e. Flow Rate at       Once/18 150 psig         months The RCIC pump shall deliver at least 600 gpm during each flow test.
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.E Hi h Pressure Coo ant In ection 4.5.E Hi h Pressure Coolant In ection 4.5.E.1 (Cont'd) e.
: 2. If the RCICS    is inoperable,                     f. Verify that          Once/Month the reactor   may   remain in                           each valve operation   for a   period not                         (manual, power-to exceed   7 days   if the                             operated, or HPCIS  is  OPERABLE   during                            automatic) in the such time.                                               injection flowpath that is not locked,
Flow Rate at Once/18 150 psig months The HPCI pump shall deliver at least 5000 gpm during each flow rate test.
: 3. If Specifications 3.5.F.1                              sealed, or other-or 3.5.F.2 are not met, an                               wise secured in orderly shutdown shall be                               position, is in its initiated and the reactor                               correct* position.
Once/Month f.
shall be depressurized to less than 150 psig within                       2. No  additional surveillances 24 hours.                                          are required.
Verify that each valve (manual, power-
* Except that an automatic valve capable of automatic return to its normal position when a signal is present may be in a position for another     mode of operation.
: operated, or automatic) in the injection flow-path that is not locked, sealed, or otherwise secured in position, is in its correct* position.
BFN Unit 1 3.5/4.5-15                AMEHDMEI'tT NO. I7 3
: 2. If the HPCI system is inoperable, the reactor may remain in operation for a period not to exceed 7 days, provided the ADS,
: CSS, RHRS (LPCI), and RCICS are OPERABLE.
2.
No additional surveillances are required.
: 3. If Specifications 3.5.E.l or 3.5.E.2 are not met, an orderly shutdown shall be initiated and the reactor vessel pressure shall be reduced to 150 psig or less within 24 hours.
Except that an automatic valve capable of automatic return to its ECCS position when an ECCS signal is present may be in a position for another mode of operation.
F.
Reactor Core Isolation Coolin F.
Reactor Core Isolation Coolin 1.
The RCICS shall be OPERABLE whenever there is irradiated fuel in the reactor vessel and the reactor vessel pressure is above 150 psig, except in the COLD SHUTDOWN CONDITION or as specified in 3.5.F.2.
OPERABILITY shall 1.
RCIC Subsystem testing shall be performed as follows:
a.
Simulated Auto-Once/18 matic Actuation months Test BFN Unit 1 3.5/4.5-14


4     CORE A D CO   AINME    COOL  G S  STEMS LIMITING CONDITIONS       FOR OPERATION             SURVEILLANCE REQUIREMENTS 3.5.G    Automatic    De  essur zation            4.5.G  Automatic  De ressurization Four  of the six valves of                  1. During each operating the Automatic                                  cycle the following Depressurization System                        tests shall be performed shall  be OPERABLE:                            on the ADS:
3 4
(1) PRIOR TO STARTUP from                     a. A simulated automatic a COLD CONDITION, or,                          actuation test shall be performed PRIOR TO (2) whenever there is                                STARTUP after each irradiated fuel in the                          refueling outage.
CORE AND CO NME COOLI G SYS EMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.F Reacto Co e
reactor vessel and the                          Manual  surveillance reactor vessel pressure                        of the relief valves is greater than 105 psig,                      is covered in except  in the  COLD SHUT-                      4.6.D.2.
so ation Coo in 4.5.F Reactor Core Iso at o
DOWN  CONDITION  or as specified in 3.5.G.2  and 3.5.G.3 below.
Coo i 3.5.F.l (Cont'd) be determined within 12 hours after reactor steam pressure reaches 150 psig from a COLD CONDITION or alternatively PRIOR TO STARTUP by using an auxiliary steam supply.
: 2. If three of the six ADS                    2. No additional surveillances valves are known to be                          are required.
4.5.F.1 (Cont'd) b.
incapable of automatic operation, the reactor may remain in operation for a period not to exceed 7 days, provided the HPCI system is OPERABLE.   (Note that the pressure relief function of these valves is assured by Section 3.6.D of these specifications and that this specification only applies to the ADS function.) If more than three of the six ADS  valves are known to be incapable of automatic operation, an immediate orderly shutdown shall be initiated, with the reactor in  a HOT SHUTDOWN CONDITION in  6 hours, and in a COLD SHUTDOWN CONDITION in the following 18 hours.
Pump OPERABILITY
: 3. If  Specifications 3.5.G.1 and 3.5.G.2 cannot be met, an orderly shutdown    will be BFN ~                                      3.5/4.5-16 Unit    1
: c. Motor-Operated Valve OPERABILITY Per Specifi-cation 1.0.MM Per Specifi-cation 1.0.MM d.
Flow Rate at Once/3 normal reactor months vessel operating pressure e.
Flow Rate at 150 psig Once/18 months 2.
If the RCICS is inoperable, the reactor may remain in operation for a period not to exceed 7 days if the HPCIS is OPERABLE during such time.
3.
If Specifications 3.5.F.1 or 3.5.F.2 are not met, an orderly shutdown shall be initiated and the reactor shall be depressurized to less than 150 psig within 24 hours.
The RCIC pump shall deliver at least 600 gpm during each flow test.
Once/Month f.
Verify that each valve (manual, power-
: operated, or automatic) in the injection flowpath that is not locked,
: sealed, or other-wise secured in position, is in its correct* position.
2.
No additional surveillances are required.
Except that an automatic valve capable of automatic return to its normal position when a signal is present may be in a position for another mode of operation.
BFN Unit 1 3.5/4.5-15 AMEHDMEI'tTNO. I7 3


5 4       CORE AND CO     A    NT COOLI G S S EMS LIMITING CONDITIONS     FOR OPERATION           SURVEILLANCE REQUIREMENTS 3.5.G     Automatic   De   ressurization          4.5.G   Automatic     De ressurizatio 3.5.G.3 (Cont'd) initiated and the reactor vessel pressure shall be reduced to 105 psig or less within 24 hours.
4 CORE A
H. Maintenance    o  F    ed Dischar e        H. Ma  ntenance    of Filled Dischar  e
D CO AINME COOL G
          ~Pi e                                          ~Pi e Whenever the core spray systems,                The  following surveillance LPCI, HPCI, or RCIC are required                requirements shall be adhered to be OPERABLE, the discharge                  to assure that the discharge piping from the pump discharge                  piping of the core spray of these systems to the last                    systems, LPCI, HPCI, and RCIC block valve shall be filled.                    are filled:
S STEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.G Automatic De essur zation 4.5.G Automatic De ressurization Four of the six valves of the Automatic Depressurization System shall be OPERABLE:
The  suction of the RCIC and HPCI              1. Every month and prior to the pumps  shall be aligned to the                       testing of the RHRS (LPCI and condensate    storage tank, and                      Containment Spray) and core the pressure suppression chamber                      spray system, the discharge head tank shall normally be                           piping of these systems shall aligned to serve the discharge                        be vented from the high point piping of the RHR and CS pumps.                       and water flow determined.
1.
The condensate head tank may be used to serve the RHR and CS                    2. Following any period where the discharge piping      if the PSC head                LPCI  or core spray systems stank is unavailable. The                              have  not been required to be pressure indicators on the                            OPERABLE, the discharge piping discharge of the RHR and CS                          of the inoperable system shall pumps  shall indicate not less                      be vented from the high point than  listed  below.                                 prior to the return of the system to service.
During each operating cycle the following tests shall be performed on the ADS:
Pl-75-20        48 psig Pl-75-48        48 psig                    3. Whenever the HPCI or RCIC Pl-74-51        48 psig                          system is lined up to take P1-74-65        48 psig                          suction from the condensate storage tank, the discharge piping of the HPCI and RCIC shall be vented from the high point of the system and water flow observed on a monthly basis.
(1) PRIOR TO STARTUP from a
: 4. When  the  RHRS and the CSS  are required to be   OPERABLE, the pressure indicators which monitor the discharge lines shall be monitored daily and the pressure recorded.
COLD CONDITION, or, (2) whenever there is irradiated fuel in the reactor vessel and the reactor vessel pressure is greater than 105 psig, except in the COLD SHUT-DOWN CONDITION or as specified in 3.5.G.2 and 3.5.G.3 below.
BFN                                        3.5/4.5-17 Unit  1
a.
A simulated automatic actuation test shall be performed PRIOR TO STARTUP after each refueling outage.
Manual surveillance of the relief valves is covered in 4.6.D.2.
: 2. If three of the six ADS valves are known to be incapable of automatic operation, the reactor may remain in operation for a period not to exceed 7 days, provided the HPCI system is OPERABLE. (Note that the pressure relief function of these valves is assured by Section 3.6.D of these specifications and that this specification only applies to the ADS function.) If more than three of the six ADS valves are known to be incapable of automatic operation, an immediate orderly shutdown shall be initiated, with the reactor in a HOT SHUTDOWN CONDITION in 6 hours, and in a COLD SHUTDOWN CONDITION in the following 18 hours.
2.
No additional surveillances are required.
: 3. If Specifications 3.5.G.1 and 3.5.G.2 cannot be met, an orderly shutdown will be BFN
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Unit 1 3.5/4.5-16


4   PRIMARY SYSTE    0    ARY LIMITING CONDITIONS FOR OPERATION                 SURVEILLANCE RE UIREMENTS 4.6.C. Coo ant  eaka e
5 4 CORE AND CO A
: 1. a. Any time  irradiated                      1. Reactor coolant fuel is in the                                 system leakage shall reactor vessel and                             be checked by the reactor coolant                                sump and air sampling temperature  is above                        system and recorded 212 F,  reactor coolant                        at least once per leakage into the                               4 hours.
NT COOLI G
primary containment from unidentified sources shall not exceed 5 gpm. In addition, the total reactor coolant system leakage into the primary containment shall not exceed 25    gpm.
S S
: b. Anytime the reactor    is in RUN MODE,   reactor coolant leakage into the primary containment from unidentified sources shall not increase by more than 2 gpm averaged over any 24-hour period in which the reactor is in the RUN MODE except as defined in 3.6.C.l.c below.
EMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.G Automatic De ressurization 4.5.G Automatic De ressurizatio 3.5.G.3 (Cont'd) initiated and the reactor vessel pressure shall be reduced to 105 psig or less within 24 hours.
: c. During the   first  24 hours in the RUN  MODE  following STARTUP, an  increase in reactor coolant leakage into the primary containment of >2 gpm is acceptable as long as the requirements of 3.6.C.l.a are met.
H.
BFN                                       3.6/4.6-9              AMENDMENT NO. I3 7 Unit  1
Maintenance o
F ed Dischar e
~Pi e
H.
Ma ntenance of Filled Dischar e
~Pi e
Whenever the core spray systems, LPCI, HPCI, or RCIC are required to be OPERABLE, the discharge piping from the pump discharge of these systems to the last block valve shall be filled.
The following surveillance requirements shall be adhered to assure that the discharge piping of the core spray
: systems, LPCI, HPCI, and RCIC are filled:
The suction of the RCIC and HPCI pumps shall be aligned to the condensate storage
: tank, and the pressure suppression chamber head tank shall normally be aligned to serve the discharge piping of the RHR and CS pumps.
The condensate head tank may be used to serve the RHR and CS discharge piping if the PSC head stank is unavailable.
The pressure indicators on the discharge of the RHR and CS pumps shall indicate not less than listed below.
1.
Every month and prior to the testing of the RHRS (LPCI and Containment Spray) and core spray system, the discharge piping of these systems shall be vented from the high point and water flow determined.
2.
Following any period where the LPCI or core spray systems have not been required to be OPERABLE, the discharge piping of the inoperable system shall be vented from the high point prior to the return of the system to service.
Pl-75-20 Pl-75-48 Pl-74-51 P1-74-65 48 psig 48 psig 48 psig 48 psig 3.
Whenever the HPCI or RCIC system is lined up to take suction from the condensate storage
: tank, the discharge piping of the HPCI and RCIC shall be vented from the high point of the system and water flow observed on a monthly basis.
4.
When the RHRS and the CSS are required to be OPERABLE, the pressure indicators which monitor the discharge lines shall be monitored daily and the pressure recorded.
BFN Unit 1 3.5/4.5-17


4   PRIMARY SYSTE     OUNDARY LIMITING CONDITIONS   FOR OPERATION               SURVEILLANCE REQUIREMENTS 3.6.C   Coolant Leaka  e                         4.6.C  Coolant Leaka  e
4 PRIMARY SYSTE 0
: 2. Both the sump and air sampling                2. With the air sampling systems shall be OPERABLE                          system inoperable, grab during  REACTOR POWER OPERATION.                  samples  shall be From and after the date that                      obtained and analyzed at one of these systems is made                      least once every 24 or found to be inoperable for                      hours.
ARY LIMITING CONDITIONS FOR OPERATION SURVEILLANCE RE UIREMENTS 4.6.C.
any reason, REACTOR POWER OPERATION  is permissible only during the succeeding 24 hours for the sump system or 72 hours for the air sampling system.
Coo ant eaka e
The air sampling system may be removed from service for a period of  4 hours for calibration, function testing, and maintenance without providing  a temporary monitor.
1.
: 3. If the condition in   1 or  2 above cannot be met, an orderly shutdown shall be initiated and the reactor shall be placed in the COLD SHUTDOWN CONDITION  within 24  hours.
a.
D. Relief Valves                                  D. Relief Valves
Any time irradiated fuel is in the reactor vessel and reactor coolant temperature is above 212 F, reactor coolant leakage into the primary containment from unidentified sources shall not exceed 5 gpm.
: 1. When more  than one  relief  valve          l. Approximately one-half is  known  to be failed, an                        of all relief valves orderly shutdown shall be                          shall  be bench-checked initiated and the reactor                          or replaced with a depressurized to less than 105                    bench-checked valve psig within 24 hours. The                          each operating cycle.
In addition, the total reactor coolant system leakage into the primary containment shall not exceed 25 gpm.
relief valves are not required                    All 13 valves will have to be OPERABLE in the COLD                        been checked or replaced SHUTDOWN  CONDITION.                              upon the completion of every second cycle.
1.
: 2. In accordance with Specification 1.0.MM, each relief valve shall be manually opened  until thermocouples and acoustic monitors downstream of the valve indicate steam is flowing from the valve.
Reactor coolant system leakage shall be checked by the sump and air sampling system and recorded at least once per 4 hours.
BFN                                     3.6/4.6-10 Unit  1
b.
Anytime the reactor is in RUN MODE, reactor coolant leakage into the primary containment from unidentified sources shall not increase by more than 2 gpm averaged over any 24-hour period in which the reactor is in the RUN MODE except as defined in 3.6.C.l.c below.
c.
During the first 24 hours in the RUN MODE following
: STARTUP, an increase in reactor coolant leakage into the primary containment of >2 gpm is acceptable as long as the requirements of 3.6.C.l.a are met.
BFN Unit 1 3.6/4.6-9 AMENDMENTNO. I3 7


3.6/4.6   BASES 3.6.C/4.6.C (Cont'd) suggest a reasonable margin of safety that such leakage magnitude would not result from a crack approaching the critical size for rapid propagation. Leakage less than the magnitude specified can be detected reasonably in a matter of a few hours utilizing the available leakage detection schemes, and     if the origin cannot be determined in a reasonably short time, the unit should be shut down to allow further investigation and corrective action.
4 PRIMARY SYSTE OUNDARY LIMITING CONDITIONS FOR OPERATION 3.6.C Coolant Leaka e
The two gpm   limit for coolant leakage rate increase over any 24 hour period is a   limit specified by the NRC (Reference 2). This limit applies only during the RUN mode to avoid being penalized for the expected coolant leakage increase during pressurization.
SURVEILLANCE REQUIREMENTS 4.6.C Coolant Leaka e
The total leakage rate consists of all leakage, identified and unidentified, which flows to the drywell floor drain and equipment drain sumps ~
2.
The capacity of the drywell floor sump pump is 50 gpm and the capacity of the drywell equipment sump pump is also 50 gpm. Removal of 25 gpm from either of these sumps can be accomplished with considerable margin.
Both the sump and air sampling systems shall be OPERABLE during REACTOR POWER OPERATION.
From and after the date that one of these systems is made or found to be inoperable for any reason, REACTOR POWER OPERATION is permissible only during the succeeding 24 hours for the sump system or 72 hours for the air sampling system.
2.
With the air sampling system inoperable, grab samples shall be obtained and analyzed at least once every 24 hours.
The air sampling system may be removed from service for a period of 4 hours for calibration, function testing, and maintenance without providing a temporary monitor.
: 3. If the condition in 1 or 2 above cannot be met, an orderly shutdown shall be initiated and the reactor shall be placed in the COLD SHUTDOWN CONDITION within 24 hours.
D.
Relief Valves D.
Relief Valves 1.
When more than one relief valve is known to be failed, an orderly shutdown shall be initiated and the reactor depressurized to less than 105 psig within 24 hours.
The relief valves are not required to be OPERABLE in the COLD SHUTDOWN CONDITION.
l.
Approximately one-half of all relief valves shall be bench-checked or replaced with a bench-checked valve each operating cycle.
All 13 valves will have been checked or replaced upon the completion of every second cycle.
2.
In accordance with Specification 1.0.MM, each relief valve shall be manually opened until thermocouples and acoustic monitors downstream of the valve indicate steam is flowing from the valve.
BFN Unit 1 3.6/4.6-10
 
3.6/4.6 BASES 3.6.C/4.6.C (Cont'd) suggest a reasonable margin of safety that such leakage magnitude would not result from a crack approaching the critical size for rapid propagation.
Leakage less than the magnitude specified can be detected reasonably in a matter of a few hours utilizing the available leakage detection
: schemes, and if the origin cannot be determined in a reasonably short time, the unit should be shut down to allow further investigation and corrective action.
The two gpm limit for coolant leakage rate increase over any 24 hour period is a limit specified by the NRC (Reference 2).
This limit applies only during the RUN mode to avoid being penalized for the expected coolant leakage increase during pressurization.
The total leakage rate consists of all leakage, identified and unidentified, which flows to the drywell floor drain and equipment drain sumps
~
The capacity of the drywell floor sump pump is 50 gpm and the capacity of the drywell equipment sump pump is also 50 gpm.
Removal of 25 gpm from either of these sumps can be accomplished with considerable margin.
REFERENCE
REFERENCE
: 1. Nuclear System Leakage Rate Limits (BFNP FSAR Subsection 4.10)
: 1. Nuclear System Leakage Rate Limits (BFNP FSAR Subsection 4.10)
: 2. Safety Evaluation Report (SER) on IE Bulletin 82-03 D   4   D Re   ef Valves To meet   the safety basis, 13   relief valves have been installed on the unit with   a total capacity of 84.1 percent of nuclear boiler rated steam flow at a reference pressure of (1,105 + 1 percent) psig. The analysis of the worst overpressure transient, (3-second closure of all main steam line isolation valves) neglecting the direct scram (valve position scram) results in a maximum vessel pressure which, if a neutron flux scram is assumed considering 12 valves operable, results in adequate margin to the code allowable overpressure limit of 1,375 psig.
: 2. Safety Evaluation Report (SER) on IE Bulletin 82-03 D 4 D
To meet   operational design, the analysis of the plant isolation transient (generator load reject with bypass valve failure to open) shows that 12 of the 13 relief valves limit peak system pressure to a value which is well below the allowed vessel overpressure of 1,375 psig.
Re ef Valves To meet the safety basis, 13 relief valves have been installed on the unit with a total capacity of 84.1 percent of nuclear boiler rated steam flow at a reference pressure of (1,105
      'Experience in relief valve operation shows that a testing of 50 percent of the valves per year is adequate to detect failures or deteriorations.
+ 1 percent) psig.
The relief valves are benchtested every second operating cycle to ensure that their setpoints are within the g 1 percent tolerance. The relief valves are tested in place in accordance with Specification 1.0.MN to establish that they will open and pass steam.
The analysis of the worst overpressure transient, (3-second closure of all main steam line isolation valves) neglecting the direct scram (valve position scram) results in a maximum vessel pressure which, if a neutron flux scram is assumed considering 12 valves operable, results in adequate margin to the code allowable overpressure limit of 1,375 psig.
BFN                                         3.6/4.6-30         AMENDMENT NO g~O Unit  1
To meet operational
: design, the analysis of the plant isolation transient (generator load reject with bypass valve failure to open) shows that 12 of the 13 relief valves limit peak system pressure to a value which is well below the allowed vessel overpressure of 1,375 psig.
'Experience in relief valve operation shows that a testing of 50 percent of the valves per year is adequate to detect failures or deteriorations.
The relief valves are benchtested every second operating cycle to ensure that their setpoints are within the g 1 percent tolerance.
The relief valves are tested in place in accordance with Specification 1.0.MN to establish that they will open and pass steam.
BFN Unit 1 3.6/4.6-30 AMENDMENT NO g~O


3.6/4.6   ~BAS S 3.6.D/4.6.D (Cont'd)
3.6/4.6
The requirements established above apply when the nuclear system can be pressurized above ambient conditions. These requirements are applicable at nuclear system pressures below normal operating pressures because abnormal operational transients could possibly start at these conditions such that eventual overpressure relief would be needed.               However, these transients are much less severe, in terms of pressure, than those starting at rated conditions. The valves need not be functional when the vessel head is removed, since the nuclear system cannot be pressurized.
~BAS S
The   relief   valves are not required to be OPERABLE in the COLD SHUTDOWN CONDITION. Overpressure protection is provided during hydrostatic tests by two of the relief valves whose relief setting has been established in conformance with ASME Section XI code requirements.             The capacity of one relief valve exceeds       the charging   capacity   of the pressurization   source used during       hydrostatic   testing. Two relief valves are used to provide redundancy.
3.6.D/4.6.D (Cont'd)
REFERENCES
The requirements established above apply when the nuclear system can be pressurized above ambient conditions.
: 1. Nuclear System Pressure Relief System         (BFNP FSAR   Subsection 4.4)
These requirements are applicable at nuclear system pressures below normal operating pressures because abnormal operational transients could possibly start at these conditions such that eventual overpressure relief would be needed.
: 2. Amendment 22   in response   to AEC Question 4.2 of December 6, 1971.
: However, these transients are much less severe, in terms of pressure, than those starting at rated conditions.
: 3.   "Protection Against Overpressure"         (ASME Boiler and Pressure   Vessel Code,   Section III, Article 9)
The valves need not be functional when the vessel head is removed, since the nuclear system cannot be pressurized.
: 4. Browns Ferry Nuclear     Plant Design Deficiency Report      Target Rock Safety-Relief Valves, transmitted by J. E. Gilliland to F. E. Kruesi, August 29, 1973
The relief valves are not required to be OPERABLE in the COLD SHUTDOWN CONDITION.
: 5. Generic Reload Fuel Application, Licensing Topical Report, NEDE-24011-P-A and Addenda 3.6.E/4.6.E     ~Jet Pum s Failure of     a jet pump nozzle assembly holddown mechanism, nozzle assembly and/or riser, would increase the cross-sectional flow area for blowdown following the design basis double-ended line break. Also, failure of the diffuser would eliminate the capability to reflood the core to two-thirds height level following a recirculation line break. Therefore,                 if a failure occurred, repairs       must   be made.
Overpressure protection is provided during hydrostatic tests by two of the relief valves whose relief setting has been established in conformance with ASME Section XI code requirements.
The   detection technique is as follows. With the two recirculation pumps balanced     in speed to within + 5 percent, the flow rates in both recirculation loops will be verified by control room monitoring instruments. If the two flow rate values do not differ by more than 10 percent, riser and nozzle assembly integrity has been verified.
The capacity of one relief valve exceeds the charging capacity of the pressurization source used during hydrostatic testing.
BFN                                           3.6/4.6-31 Unit  1
Two relief valves are used to provide redundancy.
REFERENCES 1.
Nuclear System Pressure Relief System (BFNP FSAR Subsection 4.4) 2.
Amendment 22 in response to AEC Question 4.2 of December 6, 1971.
3.
"Protection Against Overpressure" (ASME Boiler and Pressure Vessel Code, Section III, Article 9) 4.
Browns Ferry Nuclear Plant Design Deficiency ReportTarget Rock Safety-Relief Valves, transmitted by J. E. Gilliland to F. E. Kruesi, August 29, 1973 5.
Generic Reload Fuel Application, Licensing Topical Report, NEDE-24011-P-A and Addenda 3.6.E/4.6.E
~Jet Pum s Failure of a jet pump nozzle assembly holddown mechanism, nozzle assembly and/or riser, would increase the cross-sectional flow area for blowdown following the design basis double-ended line break.
Also, failure of the diffuser would eliminate the capability to reflood the core to two-thirds height level following a recirculation line break.
Therefore, if a failure occurred, repairs must be made.
The detection technique is as follows.
With the two recirculation pumps balanced in speed to within + 5 percent, the flow rates in both recirculation loops will be verified by control room monitoring instruments.
If the two flow rate values do not differ by more than 10 percent, riser and nozzle assembly integrity has been verified.
BFN Unit 1 3.6/4.6-31


3.6/4.6   BASES 3.6.E/4.6.E (Cont'd)
3.6/4.6 BASES 3.6.E/4.6.E (Cont'd)
If they   do differ by 10 percent or more, the core flow rate measured by the jet pump diffuser differential pressure     system must be checked against the core flow rate derived from the measured values of loop flow to core flow correlation.       If the difference between measured and derived core flow rate is 10 percent or more (with the derived value higher) diffuser measurements will be taken to define the location within the vessel of failed jet pump nozzle (or riser) and the unit shut down for repairs.     If the potential blowdown flow area is increased, the system resistance to the recirculation pump is also reduced; hence, the affected drive pump will "run out" to a substantially higher flow rate (approximately 115 percent to 120 percent for a single nozzle failure).
If they do differ by 10 percent or more, the core flow rate measured by the jet pump diffuser differential pressure system must be checked against the core flow rate derived from the measured values of loop flow to core flow correlation. If the difference between measured and derived core flow rate is 10 percent or more (with the derived value higher) diffuser measurements will be taken to define the location within the vessel of failed jet pump nozzle (or riser) and the unit shut down for repairs.
If the two loops are balancedcannot  in flow at the same pump speed, the resistance     characteristics           have changed. Any imbalance between drive loop flow rates would be indicated by the plant process instrumentation. In addition, the affected jet pump would provide a leakage path past the core thus reducing the core flow rate. The reverse flow through the inactive jet pump would still be indicated by a positive differential pressure but the net effect would be a slight decrease (3 percent to 6 percent) in the total core flow measured.           This decrease, together with the loop flow increase, would result in a lack of correlation between measured and derived core flow rate. Finally, the affected jet pump diffuser differential pressure signal would be reduced because the backflow would be less than the normal forward flow.
If the potential blowdown flow area is increased, the system resistance to the recirculation pump is also reduced;
A nozzle-riser system failure could also generate the coincident failure of a jet pump diffuser body; however, the converse is not true. The lack of any substantial stress in the jet pump diffuser body makes failure impossible without an initial nozzle-riser system failure.
: hence, the affected drive pump will "run out" to a substantially higher flow rate (approximately 115 percent to 120 percent for a single nozzle failure).
3.6.F/4.6.F     ecircu at on   Pum   0 e at o Steady-state     operation without forced recirculation       will not be permitted for   more than 12 hours.     And the   start of a recirculation pump   from the natural circulation condition will not be permitted unless the temperature difference between the loop to be started and the core coolant temperature is less than 75 F. This reduces the positive reactivity insertion to an acceptably low value.
If the two loops are balanced in flow at the same pump speed, the resistance characteristics cannot have changed.
Any imbalance between drive loop flow rates would be indicated by the plant process instrumentation.
In addition, the affected jet pump would provide a
leakage path past the core thus reducing the core flow rate.
The reverse flow through the inactive jet pump would still be indicated by a positive differential pressure but the net effect would be a slight decrease (3 percent to 6 percent) in the total core flow measured.
This decrease, together with the loop flow increase, would result in a lack of correlation between measured and derived core flow rate.
Finally, the affected jet pump diffuser differential pressure signal would be reduced because the backflow would be less than the normal forward flow.
A nozzle-riser system failure could also generate the coincident failure of a jet pump diffuser body; however, the converse is not true.
The lack of any substantial stress in the jet pump diffuser body makes failure impossible without an initial nozzle-riser system failure.
3.6.F/4.6.F ecircu at on Pum 0 e at o
Steady-state operation without forced recirculation will not be permitted for more than 12 hours.
And the start of a recirculation pump from the natural circulation condition will not be permitted unless the temperature difference between the loop to be started and the core coolant temperature is less than 75 F.
This reduces the positive reactivity insertion to an acceptably low value.
Requiring the discharge valve of the lower speed loop to remain closed until the speed of the faster pump is below 50% of its rated speed provides assurance when going from one-to-two pump operation that excessive vibration of the jet pump risers will not occur.
Requiring the discharge valve of the lower speed loop to remain closed until the speed of the faster pump is below 50% of its rated speed provides assurance when going from one-to-two pump operation that excessive vibration of the jet pump risers will not occur.
3.6.G/4.6.G     Structural Inte   rit The requirements     for the reactor coolant systems inservice inspection program have been identified by evaluating the need for a sampling examination of areas of high stress and highest probability of failure             in the system and the need to meet as closely as possible the requirements of Section XI, of the     ASME Boiler   and Pressure   Vessel Code.
3.6.G/4.6.G Structural Inte rit The requirements for the reactor coolant systems inservice inspection program have been identified by evaluating the need for a sampling examination of areas of high stress and highest probability of failure in the system and the need to meet as closely as possible the requirements of Section XI, of the ASME Boiler and Pressure Vessel Code.
BFN                                       3.6/4.6-32 Unit
BFN Unit 3.6/4.6-32


3.6/4.6   BASES 3.6.G/4.6.G (Cont'd)
3.6/4.6 BASES 3.6.G/4.6.G (Cont'd)
The program   reflects the built-in limitations of   access to the reactor coolant systems.
The program reflects the built-in limitations of access to the reactor coolant systems.
It is intended that the required examinations and inspection be completed during each 10-year interval. The periodic examinations are to be done during refueling outages or other extended plant shutdown periods.
It is intended that the required examinations and inspection be completed during each 10-year interval.
Only proven nondestructive testing techniques       will be used.
The periodic examinations are to be done during refueling outages or other extended plant shutdown periods.
More   frequent inspections shall be performed on certain circumferential pipe welds as   listed in Section 4.6.G.4 to provide additional protection against pipe   whip. These welds were selected in respect to their distance from hangers or supports wherein a failure of the weld would permit the unsupported segments of pipe to strike the drywell wall or nearby auxiliary systems or control systems. Selection was based on judgment from actual plant observation of hanger and support locations and review of drawings.     Inspection of all these welds during each 10-year inspection interval will result in three additional examinations above the requirements of Section XI of ASME Code.
Only proven nondestructive testing techniques will be used.
An augmented   inservice surveillance program is required to determine whether any stress corrosion has occurred in any stainless steel piping, stainless components, and highly-stressed alloy steel such as hanger springs, as a result of environmental conditions associated with the March 22, 1975   fire.
More frequent inspections shall be performed on certain circumferential pipe welds as listed in Section 4.6.G.4 to provide additional protection against pipe whip.
g~EF RELICS
These welds were selected in respect to their distance from hangers or supports wherein a failure of the weld would permit the unsupported segments of pipe to strike the drywell wall or nearby auxiliary systems or control systems.
: 1. Inservice Inspection   and Testing (BFNP FSAR Subsection 4.12)
Selection was based on judgment from actual plant observation of hanger and support locations and review of drawings.
: 2. Inservice Inspection of Nuclear Reactor Coolant Systems, Section XI, ASME Boiler and Pressure Vessel Code
Inspection of all these welds during each 10-year inspection interval will result in three additional examinations above the requirements of Section XI of ASME Code.
: 3. ASME Boiler and Pressure   Vessel Code, Section III (1968 Edition)
An augmented inservice surveillance program is required to determine whether any stress corrosion has occurred in any stainless steel piping, stainless components, and highly-stressed alloy steel such as hanger
: 4. American Society   for Nondestructive Tgsting   No. SNT-TC-1A (1968 Edition)
: springs, as a result of environmental conditions associated with the March 22, 1975 fire.
: 5. Mechanical Maintenance Instruction 46 (Mechanical Equipment, Concrete, and Structural Steel Cleaning Procedure for Residue From Plant Fire Units 1 and 2)
g~EF RELICS 1.
: 6. Mechanical Maintenance   Instruction 53 (Evaluation of Corrosion Damage of Piping Components   Which Were Exposed to Residue From March 22, 1975 Fire)
Inservice Inspection and Testing (BFNP FSAR Subsection 4.12) 2.
: 7. Plant Safety Analysis   (BFNP FSAR   Subsection 4.12)
Inservice Inspection of Nuclear Reactor Coolant Systems, Section XI, ASME Boiler and Pressure Vessel Code 3.
BFN                                     3.6/4.6-33 Unit 1
ASME Boiler and Pressure Vessel Code, Section III (1968 Edition) 4.
American Society for Nondestructive Tgsting No. SNT-TC-1A (1968 Edition) 5.
Mechanical Maintenance Instruction 46 (Mechanical Equipment,
: Concrete, and Structural Steel Cleaning Procedure for Residue From Plant Fire Units 1 and 2) 6.
Mechanical Maintenance Instruction 53 (Evaluation of Corrosion Damage of Piping Components Which Were Exposed to Residue From March 22, 1975 Fire) 7.
Plant Safety Analysis (BFNP FSAR Subsection 4.12)
BFN Unit 1 3.6/4.6-33


UNIT 2 EFFECTIVE PAGE LIST REMOVE                         INSERT 3.2/4.2-14                     3.2/4.2-14 3.2/4.2-15                      3.2/4.2-15 3.2/4.2-16                      3.2/4.2-16*
UNIT 2 EFFECTIVE PAGE LIST REMOVE INSERT 3.2/4.2-14 3.2/4.2-15 3.2/4.2-16 3.2/4.2-17 3.2/4.2-23 3.2/4.2-24 3.5/4.5-7 3.5/4.5-8 3.5/4.5-12 3.5/4.5-13 3.5/4.5-14 3.5/4.5-15 3.5/4.5-16 3.5/4.5-17 3.6/4.6-9 3.6/4.6-10 3.6/4.6-30 3.6/4.6-31 3.6/4.6-32.
3.2/4.2-17                      3.2/4.2-17 3.2/4.2-23                      3.2/4.2-23+
3.6/4.6-33 3.2/4.2-14 3.2/4.2-15 3.2/4.2-16*
3.2/4.2-24                      3.2/4.2-24 3.5/4.5-7                      3.5/4.5-7 3.5/4.5-8                      3.5/4.5-8*
3.2/4.2-17 3.2/4.2-23+
3.5/4.5-12                      3.5/4.5-12*
3.2/4.2-24 3.5/4.5-7 3.5/4.5-8*
3.5/4.5-13                      3.5/4.5-13 3.5/4.5-14                     3.5/4.5-14 3.5/4.5-15                      3.5/4.5-15*
3.5/4.5-12*
3.5/4.5-16                      3.5/4.5-16 3.5/4.5-17                      3.5/4.5-17*
3.5/4.5-13 3.5/4.5-14 3.5/4.5-15*
3.6/4.6-9                      3.6/4.6-9+
3.5/4.5-16 3.5/4.5-17*
3.6/4.6-10                      3.6/4.6-10 3.6/4.6-30                      3.6/4.6-30*
3.6/4.6-9+
3.6/4.6-31                      3.6/4.6-31 3.6/4.6-32.                    3.6/4.6-32*
3.6/4.6-10 3.6/4.6-30*
3.6/4.6-33                      3.6/4.6-33*
3.6/4.6-31 3.6/4.6-32*
*Denotes overleaf or spillover page.
3.6/4.6-33*
*Denotes overleaf or spillover page.


TABLE 3.2.B INSTRUMENTATION THAT INITIATES OR CONTROLS THE CORE AND CONTAINMENT COOLING SYSTEMS Minimum No.
TABLE 3.2.B INSTRUMENTATION THAT INITIATES OR CONTROLS THE CORE AND CONTAINMENT COOLING SYSTEMS Minimum No.
~Tri '~l Operable Per Fun ti on              Tri L vel e   in         Action                     R mark Instrument Channel           > 470u above vessel  zero. A              1. Below  trip setting initiates Reactor Low Water Level                                                         HPCI.
Operable Per
(LIS-3-58A-D)
~Tri '~l Fun tion Tri L vel e
Instrument Channel           > 470" above vessel  zero. A              1. Multiplier relays initiate Reactor Low Water Level                                                         RCIC.
in Action R mark 2
(LIS-3-58A-D) 2              Instrument Channel           > 378" above vessel zero. A               1. Below   trip setting initiates Reactor Low Water Level                                                          CSS.
2(16)
(LS-3-58A-0)
Instrument Channel Reactor Low Water Level (LIS-3-58A-D)
Instrument Channel Reactor Low Water Level (LIS-3-58A-D)
Instrument Channel Reactor Low Water Level (LS-3-58A-0)
Instrument Channel-Reactor Low Water Level (LS-3-58A-D)
> 470u above vessel zero.
A
> 470" above vessel zero.
A
> 378" above vessel zero.
A
> 378" above vessel zero.
A 1.
Below trip setting initiates HPCI.
1.
Multiplier relays initiate RCIC.
1.
Below trip setting initiates CSS.
Hultiplier relays initiate LPCI.
Hultiplier relays initiate LPCI.
: 2. Multiplier relay from CSS initiates accident signal (15).
2.
2(16)          Instrument Channel-          > 378" above vessel  zero. A              l. Below   trip settings, in Reactor Low Water Level                                                          conjunction with drywell, (LS-3-58A-D)                                                                    high pressure. low water level permissive, 105 sec.
Multiplier relay from CSS initiates accident signal (15).
l.
Below trip settings, in conjunction with drywell, high pressure.
low water level permissive, 105 sec.
delay timer and CSS or RHR pump running, initiates ADS.
delay timer and CSS or RHR pump running, initiates ADS.
: 2. Below   trip settings, in conjunction with low reactor water level permissive, 105 sec. delay timer, 12 1/2 min. delay timer, CSS or RHR pump running, initiates   ADS.
2.
1(16)           Instrument Channel           > 544" above vessel  zero. A              1. Below  trip setting  permissive Reactor Low Water Level                                                         for initiating signals    on ADS.
Below trip settings, in conjunction with low reactor water level permissive, 105 sec.
Permissive (LIS-3-184, 185)
delay timer, 12 1/2 min. delay timer, CSS or RHR pump running, initiates ADS.
Instrument Channel-           > 312 5/16" above vessel zero. A               1. Below  trip setting  prevents Reactor Low Water Level      (2/3 core height)                                 inadvertent operation of (LIS-3-52 and LIS-3-62A)                                                        containment spray during accident condition.
1(16)
The automatic initiation capability of this instrument channel is not required to be OPERABLE while the Reactor Vessel water level monitoring modification is being performed. Hanual initiation capability of the associated system will be available during that time the automatic initiation logic is out-of-service.
Instrument Channel Reactor Low Water Level Permissive (LIS-3-184, 185)
Instrument Channel-Reactor Low Water Level (LIS-3-52 and LIS-3-62A)
> 544" above vessel zero.
A
> 312 5/16" above vessel zero.
A (2/3 core height) 1.
Below trip setting permissive for initiating signals on ADS.
1.
Below trip setting prevents inadvertent operation of containment spray during accident condition.
The automatic initiation capability of this instrument channel is not required to be OPERABLE while the Reactor Vessel water level monitoring modification is being performed.
Hanual initiation capability of the associated system will be available during that time the automatic initiation logic is out-of-service.


                                                                                                                  ~
~
                                                                                                                    ~
~
TABLE 3.2.8 (Continued)
TABLE 3.2.8 (Continued)
Hinimum No.
Hinimum No.
Operable Per
Operable Per T~ri S
                                                                  ~Ainn T~ri S   i 2(18)
i 2(18) 2(18)
Func i n Instrument Channel-Drywell High Pressure 1<
Func i n Instrument Channel-Drywell High Pressure (PIS-64-58 E-H)
Tri L v 1 p<2.5 psig in A'emarks
Instrument Channel-Drywell High Pressure (PIS-64-58 A-D)
: l. Below   trip setting prevents inadvertent operation of (PIS-64-58 E-H)                                            containment spray during accident conditions.
Tri L v 1
2(18)      Instrument Channel-  <  2.5 psig                      l. Above   trip setting in con-Drywell High Pressure                                      junction with low reactor (PIS-64-58 A-D)                                            pressure initiates CSS.
1< p<2.5 psig
< 2.5 psig in
~AinnA'emarks l.
Below trip setting prevents inadvertent operation of containment spray during accident conditions.
l.
Above trip setting in con-junction with low reactor pressure initiates CSS.
Hultiplier relays initiate HPCI.
Hultiplier relays initiate HPCI.
: 2. Hul tiplier relay    from CSS ini ti ates ace i dent  s i gnal . (15) 2(18)     Instrument Channel-   <  2.5 psig                      l. Above  trip setting in Drywell High Pressure                                     conjunction with low (PIS-64-58A-D)                                             reactor pressure initiates LPCI.
2(18) 2(16) (18)
2(16) (18) Instrument Channel-   < 2.5 psig                       1. Above   trip setting, in Drywell High Pressure                                      conjunction with low reactor (PIS-64-57A-D)                                            water level, low reactor water level permissive, 105 sec. delay timer and CSS or RHR pump running, initiates   ADS.
Instrument Channel-Drywell High Pressure (PIS-64-58A-D)
Instrument Channel-Drywell High Pressure (PIS-64-57A-D)
< 2.5 psig
< 2.5 psig 2.
Hultiplier relay from CSS initiates ace ident s ignal. (15) l.
Above trip setting in conjunction with low reactor pressure initiates LPCI.
1.
Above trip setting, in conjunction with low reactor water level, low reactor water level permissive, 105 sec.
delay timer and CSS or RHR pump running, initiates ADS.


                                                                                                            ~ ~
~
TABLE   3.2.B (Continued)
~
TABLE 3.2.B (Continued)
Hinimum No.
Hinimum No.
Operable Per
Operable Per
~Tri S   1         Fun ti n           Tri  L  vel Set in        A~ion            Remarks Instrument Channel         450  psig    + 15                1. Below  trip setting permissive Reactor Low Pressure                                           for opening CSS and LPCI (PIS-3-74 A 8 B)                                               admission valves.
~Tri S
1 Fun ti n Instrument Channel Reactor Low Pressure (PIS-3-74 A 8 B)
(PIS-68-95, 96)
(PIS-68-95, 96)
Instrument Channel-         230  psig +    15              1. Recirculation discharge valve Reactor Low Pressure                                           actuation.
Instrument Channel-Reactor Low Pressure (PS-3-74 A 5 B)
(PS-3-74 A 5 B)
(PS-68-95, 96)
(PS-68-95, 96)
Instrument Channel         100  psig  + 15                l. Below  trip setting in Reactor Low Pressure                                           conjunction with (PS-68-93 8 94, SW 01)                                         containment isolation signal and both suction valves open will close RHR (LPCI) admission valves.
Instrument Channel Reactor Low Pressure (PS-68-93 8 94, SW 01)
Core Spray Auto Sequencing 6< t   <8 sec.                 1. With diesel power Timers (5)                                                  2. One per motor LPCI Auto Sequencing       0< t   <1 sec.                 1. With diesel power Timers (5)                                                  2. One per motor RHRSW Al, B3, Cl, and 03 13<   t <15 sec.               1. With diesel power Timers                                                      2. One per pump Core Spray and LPCI Auto     0<   t <1 sec.               1. With normal power Sequencing Timers (6)        6<   t <8 sec.                 2. One per CSS motor 12<   t <16 sec.               3. Two per RHR motor 18<   t <24 sec.
Tri L vel Set in 450 psig + 15 230 psig + 15 100 psig + 15 A~ion Remarks 1.
RHRSW  Al,  B3, Cl, and 03  27<   t < 29 sec.             1. With normal power Timers                                                      2. One per pump
Below trip setting permissive for opening CSS and LPCI admission valves.
1.
Recirculation discharge valve actuation.
l.
Below trip setting in conjunction with containment isolation signal and both suction valves open will close RHR (LPCI) admission valves.
Core Spray Auto Sequencing 6< t <8 sec.
Timers (5) 1.
With diesel power 2.
One per motor LPCI Auto Sequencing Timers (5) 0< t <1 sec.
1.
With diesel power 2.
One per motor RHRSW Al, B3, Cl, and 03 13< t <15 sec.
Timers 1.
With diesel power 2.
One per pump Core Spray and LPCI Auto Sequencing Timers (6)
RHRSW Al, B3, Cl, and 03 Timers 0< t <1 sec.
6< t <8 sec.
12< t <16 sec.
18< t <24 sec.
27< t < 29 sec.
1.
With normal power 2.
One per CSS motor 3.
Two per RHR motor 1.
With normal power 2.
One per pump


TABLE 3.2.B (Continued)
TABLE 3.2.B (Continued)
Minimum No.
Minimum No.
Operable Per Ir i~!Lbll           Function               Tri L v 1     tin     ~AI II         R marks 1(16)  AOS  Timer                    105 sec + 7                        1. Above   trip setting in conjunction with low reactor water level permissive, low reactor water level, high drywell pressure or high drywell pressure bypass timer timed out, and RHR or CSS pumps running, initiates ADS.
Operable Per Iri~!Lbll 1(16)
1(16)   ADS Timer (12 1/2 min.)     12 1/2 min. +   2                 l. Above   trip setting, in (High Drywell Pressure                                              conjunction with low Bypass Timer)                                                      reactor water level permissive, low reactor water level, 105 sec.
Function AOS Timer Tri L v 1
delay timer, and RHR or CSS pumps running, initiates   AOS.
tin 105 sec
Instrument Channel-          100 +10  psig                      l. Below   trip setting   defers ADS RHR  Discharge Pressure                                            actuation.
+ 7
Instrument Channel            185 +10  psig                      l. Below   trip setting   defers AOS CSS Pump  Discharge Pressure                                      actuation.
~AI II R marks 1.
1(3)  Core Spray Sparger to        2  psid +0.4                          Alarm to detect core sparger Reactor Pressure Vessel d/p                                        pipe break.
Above trip setting in conjunction with low reactor water level permissive, low reactor water level, high drywell pressure or high drywell pressure bypass timer timed out, and RHR or CSS pumps running, initiates ADS.
RHR  (LPCI) Trip System bus  N/A                                1. Monitors   availability of power monitor Core Spray Trip System bus power monitor N/A                                1.
1(16) 1(3)
power to Moni tors power logic systems.
Instrument Channel-RHR Discharge Pressure 100 +10 psig Instrument Channel 185 +10 psig CSS Pump Discharge Pressure Core Spray Sparger to 2 psid +0.4 Reactor Pressure Vessel d/p RHR (LPCI) Trip System bus N/A power monitor Core Spray Trip System bus N/A power monitor ADS Trip System bus power N/A monitor ADS Timer (12 1/2 min.)
avail abil i ty of to logic systems.
12 1/2 min. + 2 (High Drywell Pressure Bypass Timer) l.
I
Above trip setting, in conjunction with low reactor water level permissive, low reactor water level, 105 sec.
                                                                                                                    ~
delay timer, and RHR or CSS pumps running, initiates AOS.
ADS  Trip  System bus power  N/A                                1. Monitors   availability of monitor                                                            power   to logic systems and valves.
l.
Below trip setting defers ADS actuation.
l.
Below trip setting defers AOS actuation.
Alarm to detect core sparger pipe break.
1.
Monitors availability of power to logic systems.
1.
Monitors avail abil i ty of power to logic systems.
1.
Monitors availability of power to logic systems and valves.
I~


NOTES FOR TABLE       2 B
NOTES FOR TABLE 2 B 1.
: 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.
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.
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:
Action:
A. Repair in 24 hours.       If the function is not   OPERABLE in 24 hours, take action B.
A.
B. Declare the system or component inoperable.
Repair in 24 hours. If the function is not OPERABLE in 24 hours, take action B.
C. Immediately take action       B until power   is verified on the trip system.
B.
D. No action required; indicators are considered redundant.
Declare the system or component inoperable.
: 2. In only   one trip system.
C.
: 3. Not considered     in a trip system.
Immediately take action B until power is verified on the trip system.
: 4. Requires one channel from each physical location (there are 4 locations) in the steam line space.
D.
: 5. With diesel power, each     RHRS pump   is scheduled to start immediately     and each CSS   pump is sequenced   to start about   7 sec. later.
No action required; indicators are considered redundant.
: 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.,
2.
at which time the     full complement   of CSS and RHRS pumps would be operating.
In only one trip system.
: 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 line break.
3.
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.
Not considered in a trip system.
: 8. Note 1 does not apply to       this item.
4.
: 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.
Requires one channel from each physical location (there are 4 locations) in the steam line space.
BFN                                             3.2/4.2-23 Unit 2
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 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 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 2 3.2/4.2-23


NOTES   OR TABLE     2 B   Cont'd)
NOTES OR TABLE 2
: 10. Only one   trip system for each cooler fan.
B Cont'd) 10.
: 11. In only two of the four 4160-V shutdown boards.     See note 13.
Only one trip system for each cooler fan.
: 12. In only one of the four   4160-V shutdown boards. See note 13.
11.
: 13. An emergency 4160-V shutdown board     is considered a trip system.
In only two of the four 4160-V shutdown boards.
: 14. RHRSW pump   would be inoperable. Refer to Section 4.5.C       for the requirements of a RHRSW pump being inoperable.
See note 13.
: 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.
12.
: 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.
In only one of the four 4160-V shutdown boards.
: 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 reduction shall be initiated for more than 72 hours, an and reactor power shall be orderly less than power 30 percent within four hours.
See note 13.
: 18. Not required to be     OPERABLE in the COLD SHUTDOWN CONDITION.
13.
BFN                                       3.2/4.2-24 Unit  2
An emergency 4160-V shutdown board is considered a trip system.
: 4. CORE A D CO    AINMENT COOLI    G                SYSTEMS LIMITING CONDITIONS    FOR OPERATION                              SURVEILLANCE REQUIREMENTS 3.5.B  Residual Heat    Remova  S  ste                          4.5.B  Res dua    Heat Remova S stem
14.
        ~RHRS    (LPCI and Containment                                    ~RHRS    (LPCI and Containment Cooling)                                                          Cooling)
RHRSW pump would be inoperable.
: 8. If Specifications    3.5.B.l                                8. No  additional surveillance through 3.5.B.7 are not met,                                      required.
Refer to Section 4.5.C for the requirements of a RHRSW pump being inoperable.
an  orderly shutdown shall                    be initiated  and the reactor shall be placed in the COLD SHUTDOWN CONDITION within  24  hours.
15.
: 9. When  the reactor vessel                                    9. When  the reactor vessel pressure is atmospheric and                                      pressure is atmospheric, irradiated fuel is in the                                        the RHR pumps and valves reactor vessel, at least one                                      that are required to be RHR  loop with two pumps or two                                  OPERABLE  shall be loops with one pump per loop                                      demonstrated to be OPERABLE shall  be OPERABLE. The                                      per Specification 1.0.MM.
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.
diesel generators pumps'ssociated must also be OPERABLE.
16.
: 10. If the conditions of                                        10. No  additional surveillance Specification 3.5.A.5 are met,                                    required.
The ADS circuitry is capable of accomplishing its protective action with one OPERABLE trip system.
LPCI and containment      cooling are not required.
Therefore, one trip system may be taken out of service for functional testing and calibration for a period not to exceed eight hours.
ll. When  there is irradiated fuel                              11. The RHR pumps on the adjacent units which supply in the reactor and the reactor is not in the COLD SHUTDOWN                                      cross-connect capability CONDITION, 2 RHR pumps and                                        shall be demonstrated to be associated heat exchangers and                                    OPERABLE per Specification valves on an adjacent unit                                        1.0.MM when the cross-must be OPERABLE and capable                                      connect  capability of supplying cross-connect                                        is required.
17.
capability except as specified in Specification 3.5.B.12 below. (Note:
Two RPT systems exist, either of which will trip both recirculation pumps.
Because  cross-connect capability is not a short-term requirement, a component is not considered inoperable if  cross-connect capability can be restored to service within 5 hours.)
The systems will be individually functionally tested monthly. If the test period for one RPT system exceeds two consecutive
BFN                                        3.5/4.5-7 Unit  2
: hours, the system will be declared inoperable.
If both RPT systems are inoperable or if one RPT system is inoperable for more than 72 hours, 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 2 3.2/4.2-24


4     CORE AND CO               COOLING SYSTEMS LIMITING CONDITIONS     FOR OPERATION                 SURVEILLANCE REQUIREMENTS 3.5.B     esidual Heat     e oval  S ste             4.5.B   es dua   eat Removal S stem
4.
        ~RHRS   (LPCI and Containment                        PHRS (LPCI   and Containment Cooling)                                            Cooling)
CORE A D CO AINMENT COOLI G SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.B Residual Heat Remova S ste
: 12. If three  RHR pumps  or associated          12. No additional surveillance required.
~RHRS (LPCI and Containment Cooling) 4.5.B Res dua Heat Remova S stem
heat exchangers    located on the unit cross-connection in the adjacent units are inoperable for any reason (including valve inoperability, pipe break, etc.), the reactor may remain in operation for a period not to exceed 30 days provided the remaining RHR pump and associated diesel generator are OPERABLE.
~RHRS (LPCI and Containment Cooling)
: 13. If RHR cross-connection    flow or            13. No additional surveillance heat removal capability is lost,                    required.
: 8. If Specifications 3.5.B.l through 3.5.B.7 are not met, an orderly shutdown shall be initiated and the reactor shall be placed in the COLD SHUTDOWN CONDITION within 24 hours.
the unit may remain in operation for a period not to exceed 10 days unless such capability is restored.
8.
: 14. All recirculation    pump                    14. All recirculation  pump discharge valves shall                              discharge valves shall be OPERABLE PRIOR    TO                            be tested for OPERABILITY STARTUP  (or closed  if                          during any period of permitted elsewhere                                COLD SHUTDOWN CONDITION in these specifications).                           exceeding 48 hours,  if OPERABILITY tests have not been performed during the preceding 31 days.
No additional surveillance required.
AMENDMENT NO. X6  9 BFN                                             3.5/4.5-8 Unit  2
9.
When the reactor vessel pressure is atmospheric and irradiated fuel is in the reactor vessel, at least one RHR loop with two pumps or two loops with one pump per loop shall be OPERABLE.
The pumps'ssociated diesel generators must also be OPERABLE.
9.
When the reactor vessel pressure is atmospheric, the RHR pumps and valves that are required to be OPERABLE shall be demonstrated to be OPERABLE per Specification 1.0.MM.
: 10. If the conditions of Specification 3.5.A.5 are met, LPCI and containment cooling are not required.
10.
No additional surveillance required.
ll.
When there is irradiated fuel in the reactor and the reactor is not in the COLD SHUTDOWN CONDITION, 2 RHR pumps and associated heat exchangers and valves on an adjacent unit must be OPERABLE and capable of supplying cross-connect capability except as specified in Specification 3.5.B.12 below.
(Note:
Because cross-connect capability is not a short-term requirement, a component is not considered inoperable if cross-connect capability can be restored to service within 5 hours.)
11.
The RHR pumps on the adjacent units which supply cross-connect capability shall be demonstrated to be OPERABLE per Specification 1.0.MM when the cross-connect capability is required.
BFN Unit 2 3.5/4.5-7


4     CORE AND CO             COOLI    G  SYSTEMS LIMITING CONDITIONS       FOR OPERATION             SURVEILLANCE REQUIREMENTS 3.5.C    RHR  Service Water and Emer enc            4.5.C  RHR Service Water and Emer enc E  ui  ment Coo in Wate      S  stems                u  ment, Coolin Water S stems EECWS    Cont nued                              EECWS    Continued
4 CORE AND CO COOLING SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.B esidual Heat e oval S ste
: 4. Three  of the Dl, D2,   Bl, B2            4. No additional surveillance RHRSW pumps assigned    to  the                  is required.
~RHRS (LPCI and Containment Cooling)
RHR heat exchanger supplying the standby coolant supply connection may be inoperable for a period not to exceed 30 days provided the OPERABLE pump is aligned to supply the RHR heat exchanger header and the associated diesel generator and essential control valves are OPERABLE.
: 12. If three RHR pumps or associated heat exchangers located on the unit cross-connection in the adjacent units are inoperable for any reason (including valve inoperability, pipe break, etc.),
: 5. The standby  coolant supply capability may be inoperable for a period not to exceed 10 days.
the reactor may remain in operation for a period not to exceed 30 days provided the remaining RHR pump and associated diesel generator are OPERABLE.
: 6. If Specifications  3.5.C.2 through 3.5.C.5 are not met, an orderly shutdown shall be initiated and the unit placed in the COLD SHUTDOWN CONDITION  within 24  hours.
: 13. If RHR cross-connection flow or heat removal capability is lost, the unit may remain in operation for a period not to exceed 10 days unless such capability is restored.
: 7. There  shall be at least 2 RHRSW pumps,   associated with the selected RHR    pumps, aligned for RHR heat exchanger service for each reactor vessel containing irradiated fuel.
4.5.B es dua eat Removal S stem PHRS (LPCI and Containment Cooling) 12.
BFN                                             3.5/4.5-12            AMfNDMEHTNO. I6 9 Unit  2
No additional surveillance required.
13.
No additional surveillance required.
14.
All recirculation pump discharge valves shall be OPERABLE PRIOR TO STARTUP (or closed if permitted elsewhere in these specifications).
14.
All recirculation pump discharge valves shall be tested for OPERABILITY during any period of COLD SHUTDOWN CONDITION exceeding 48 hours, if OPERABILITY tests have not been performed during the preceding 31 days.
BFN Unit 2 3.5/4.5-8 AMENDMENTNO. X 6 9


5      CORE     CO           COOLI G S STEMS LIMITING CONDITIONS   FOR OPERATION             SURVEILLANCE REQUIREMENTS 3.5.D          ent A ea Coo ers                4.5.D      ui  ment Area Coole s
4 CORE AND CO COOLI G SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.C RHR Service Water and Emer enc E ui ment Coo in Wate S stems EECWS Cont nued 4.5.C RHR Service Water and Emer enc u
: 1. The equipment area    cooler                l. Each equipment area cooler associated with each RHR                           is operated in conjunction pump and the equipment                            with the equipment served area cooler associated                             by that particular cooler; with each set of core                              therefore, the equipment spray pumps (A and C                              area coolers are tested at or B and D) must be                                the same frequency as the OPERABLE at all times                              pumps which they serve.
ment, Coolin Water S stems EECWS Continued 4.
when the pump or pumps served by that specific cooler is considered to be OPERABLE.
Three of the Dl, D2, Bl, B2 RHRSW pumps assigned to the RHR heat exchanger supplying the standby coolant supply connection may be inoperable for a period not to exceed 30 days provided the OPERABLE pump is aligned to supply the RHR heat exchanger header and the associated diesel generator and essential control valves are OPERABLE.
: 2. When an equipment area cooler is not OPERABLE, the pump(s) served by that cooler must be considered inoperable for Technical Specification purposes.
4.
E. Hi h Pressure  Coolant In ect  o            E. Hi h Pressure    Coolant In 'ection  S stem  HPCIS The HPCI system  shall  be                  1. HPCI Subsystem  testing OPERABLE  whenever there  is                    shall be performed as irradiated fuel in the                            follows:
No additional surveillance is required.
reactor vessel and the reactor vessel pressure                      a. Simulated      Once/18 is greater than 150 psig,                          Automatic      months except  in the COLD SHUTDOWN                     Actuation CONDITION   or as specified                        Test in 3.5.E.2. OPERABILITY shall be determined                          b. Pump            Per within 12 hours after                              OPERA-          Specification reactor steam pressure                            BILITY          1.0.MM reaches 150 psig from a COLD CONDITION, or alter-                    c. Motor Oper-    Per natively  PRIOR TO STARTUP                        ated Valve      Specification by using an auxiliary steam                        OPERABILITY    1.0.MM supply.
5.
: d. Flow Rate at   Once/3 normal          months reactor vessel operating pressure BFN                                     3.5/4.5-13 Unit  2
The standby coolant supply capability may be inoperable for a period not to exceed 10 days.
: 6. If Specifications 3.5.C.2 through 3.5.C.5 are not
: met, an orderly shutdown shall be initiated and the unit placed in the COLD SHUTDOWN CONDITION within 24 hours.
7.
There shall be at least 2
RHRSW pumps, associated with the selected RHR pumps, aligned for RHR heat exchanger service for each reactor vessel containing irradiated fuel.
BFN Unit 2 3.5/4.5-12 AMfNDMEHTNO. I 6 9


4.5 CORE AND CO   AINME    COOLING SYS E    S LIMITING CONDITIONS   FOR OPERATION           SURVEILLANCE REQUIREMENTS 3.5.E    Hi  Pressure  Coo ant In ection    4.5.E    Hi h Pressure    Coolant In ection 4.5.E.l (Cont'd)
5 CORE CO COOLI G
: e. Flow Rate at    Once/18 150  psig      months The HPCI pump shall deliver at least 5000 gpm during each flow rate test.
S STEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.D ent A ea Coo ers 4.5.D ui ment Area Coole s 1.
: f. Verify that        Once/Month each valve (manual, power-operated, or automatic) in the injection flow-path that is not locked, sealed, or otherwise secured in position, is in its correct* position.
The equipment area cooler associated with each RHR pump and the equipment area cooler associated with each set of core spray pumps (A and C
: 2. If the HPCI system  is                2. No  additional surveillances inoperable, the reactor                    are required.
or B and D) must be OPERABLE at all times when the pump or pumps served by that specific cooler is considered to be OPERABLE.
may remain in operation for a period not to exceed 7 days, provided the ADS, CSS,  RHRS(LPCI), and RCICS are  OPERABLE.
l.
: 3. If Specifications    3.5.E.l
Each equipment area cooler is operated in conjunction with the equipment served by that particular cooler; therefore, the equipment area coolers are tested at the same frequency as the pumps which they serve.
* Except that    an automatic or 3.5.E.2 are not met,                       valve capable of automatic an  orderly shutdown shall                    return to its ECCS position be  initiated and the                         when an ECCS signal    is reactor'essel pressure                        present may be in a shall be reduced to 150                        position for another    mode psig or less within 24                        of operation.
2.
hours.
When an equipment area cooler is not OPERABLE, the pump(s) served by that cooler must be considered inoperable for Technical Specification purposes.
F. Reactor Core  Isolation Coolin            F. Reactor Core    Isolation Coolin
E.
: 1. The RCICS  shall be OPERABLE             1. RCIC Subsystem    testing shall whenever there   is irradiated                 be performed as    follows:
Hi h Pressure Coolant In ect o
fuel in the reactor vessel and the   reactor vessel                       a. Simulated Auto- Once/18 pressure is above 150 psig,                         matic Actuation months except in the COLD SHUTDOWN                         Test CONDITION or as specified in 3.5.F.2. OPERABILITY shall BFN                                     3.5/4.5-14 Unit  2
E.
Hi h Pressure Coolant In 'ection S stem HPCIS The HPCI system shall be OPERABLE whenever there is irradiated fuel in the reactor vessel and the reactor vessel pressure is greater than 150 psig, except in the COLD SHUTDOWN CONDITION or as specified in 3.5.E.2.
OPERABILITY shall be determined within 12 hours after reactor steam pressure reaches 150 psig from a COLD CONDITION, or alter-natively PRIOR TO STARTUP by using an auxiliary steam supply.
a.
Simulated Automatic Actuation Test Once/18 months b.
Pump OPERA-BILITY Per Specification 1.0.MM c.
Motor Oper-ated Valve OPERABILITY Per Specification 1.0.MM 1.
HPCI Subsystem testing shall be performed as follows:
d.
Flow Rate at normal reactor vessel operating pressure Once/3 months BFN Unit 2 3.5/4.5-13


4     CORE A    CO             COOLI  G SYSTEMS LIMITING CONDITIONS     FOR OPERATION               SURVEILLANCE REQUIREMENTS 3.5.F. Reactor    Co e  Iso ation Coolin            4.5.F Reactor Core Isolation Coolin 3.5.F.l (Cont'd)                               4.5.F.1    (Cont'd) be determined    within  12 hours            b. Pump              Per after reactor    steam pressure                  OPERABILITY      Specifi-reaches 150 psig from a      COLD                                  cation CONDITION or alternatively                                          1.0.MM PRIOR TO STARTUP by using an auxiliary  steam supply..                    c. Motor-Operated    Per Valve            Specifi-OPERABILITY      cation 1.0.MM
4.5 CORE AND CO AINME COOLING SYS E
: d. Flow Rate at     Once/3 normal reactor  months vessel operating pressure
S LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.E Hi Pressure Coo ant In ection 4.5.E Hi h Pressure Coolant In ection 4.5.E.l (Cont'd) e.
: e. Flow Rate at    Once/18 150 psig       months The RCIC pump shall deliver at least 600 gpm during each flow test.
Flow Rate at Once/18 150 psig months The HPCI pump shall deliver at least 5000 gpm during each flow rate test.
: 2. If the   RCICS  is inoperable,               f. Verify that        Once/Month the reactor   may remain in                     each valve operation   for a   period not                     (manual, power-to exceed   7 days   if the                       operated, or HPCIS  is OPERABLE   during                      automatic) in the such time.                                         injection flowpath that is not locked,
Once/Month f.
: 3. If Specifications     3.5.F.1                      sealed, or other-or 3.5.F.2 are not met, an                         wise secured in orderly shutdown shall be                         position, is in its initiated and the reactor                         correct* position.
Verify that each valve (manual, power-
shall be depressurized to less than 150 psig within                 2. No  additional surveillances 24 hours.                                    are required.
: operated, or automatic) in the injection flow-path that is not locked, sealed, or otherwise secured in position, is in its correct* position.
* Except that   an automatic valve capable of automatic return to its normal position when a signal is present may be in a position for another   mode of operation.
: 2. If the HPCI system is inoperable, the reactor may remain in operation for a period not to exceed 7 days, provided the ADS,
AMENOMBlT WO. X V 6 BFN                                           3.5/4. 5-15 Unit  2
: CSS, RHRS(LPCI),
and RCICS are OPERABLE.
2.
No additional surveillances are required.
: 3. If Specifications 3.5.E.l or 3.5.E.2 are not met, an orderly shutdown shall be initiated and the reactor'essel pressure shall be reduced to 150 psig or less within 24 hours.
Except that an automatic valve capable of automatic return to its ECCS position when an ECCS signal is present may be in a position for another mode of operation.
F.
Reactor Core Isolation Coolin F.
Reactor Core Isolation Coolin 1.
The RCICS shall be OPERABLE whenever there is irradiated fuel in the reactor vessel and the reactor vessel pressure is above 150 psig, except in the COLD SHUTDOWN CONDITION or as specified in 3.5.F.2.
OPERABILITY shall 1.
RCIC Subsystem testing shall be performed as follows:
a.
Simulated Auto-Once/18 matic Actuation months Test BFN Unit 2 3.5/4.5-14


4   CORE AND CONTA NMENT COOLING SYSTEMS LIMITING CONDITIONS     FOR OPERATION             SURVEILLANCE REQUIREMENTS 3.5.G    Automat c  De  ressu    atio            4.5.G    utomatic  De  ressurizatio
4 CORE A
: 1. Four of the six valves    of                1. During each operating the Automatic                                    cycle the following Depressurization System                          tests shall be performed shall  be OPERABLE:                              on the ADS:
CO COOLI G SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.F.
(1) PRIOR TO STARTUP from                       a. A simulated automatic a COLD CONDITION, or,                            actuation test shall be performed PRIOR   TO (2) .whenever there is                                STARTUP after each irradiated fuel in the                            refueling outage.
Reactor Co e Iso ation Coolin 4.5.F Reactor Core Isolation Coolin 3.5.F.l (Cont'd) be determined within 12 hours after reactor steam pressure reaches 150 psig from a COLD CONDITION or alternatively PRIOR TO STARTUP by using an auxiliary steam supply..
reactor vessel and the                            Manual  surveillance reactor vessel pressure                          of the relief valves is greater than 105 psig,                        is covered in except  in the  COLD SHUT-                      4.6.D.2.
4.5.F.1 (Cont'd) b.
DOWN  CONDITION  or as specified in 3.5.G.2 and 3.5.G.3 below.
Pump OPERABILITY
: 2. If three of the six ADS                      2. No  additional surveillances valves are known to be                            are required.
: c. Motor-Operated Valve OPERABILITY Per Specifi-cation 1.0.MM Per Specifi-cation 1.0.MM d.
incapable of automatic operation, the reactor may remain in operation for a period not to exceed 7 .days, provided the HPCI system is OPERABLE.    (Note that the pressure relief function of these valves is assured by Section 3.6.D of these specifications and that this specification only applies to the    ADS function.) If more than three of the six ADS valves are known to be incapable of automatic operation, an immediate orderly shutdown shall be initiated, with the reactor in  a HOT SHUTDOWN CONDITION in  6 hours, and in a COLD SHUTDOWN CONDITION in the following 18 hours.
Flow Rate at Once/3 normal reactor months vessel operating pressure e.
: 3. If Specifications     3.5.G.l and 3.5.G.2 cannot be met, an orderly shutdown     will  be initiated and the reactor BFN                                       3.5/4.5-16 Unit  2
Flow Rate at Once/18 150 psig months The RCIC pump shall deliver at least 600 gpm during each flow test.
: 2. If the RCICS is inoperable, the reactor may remain in operation for a period not to exceed 7 days if the HPCIS is OPERABLE during such time.
: 3. If Specifications 3.5.F.1 or 3.5.F.2 are not met, an orderly shutdown shall be initiated and the reactor shall be depressurized to less than 150 psig within 24 hours.
Once/Month f.
Verify that each valve (manual, power-
: operated, or automatic) in the injection flowpath that is not locked,
: sealed, or other-wise secured in position, is in its correct* position.
2.
No additional surveillances are required.
Except that an automatic valve capable of automatic return to its normal position when a signal is present may be in a position for another mode of operation.
BFN Unit 2 3.5/4. 5-15 AMENOMBlTWO. X V 6


LIMITING CONDITIONS       FOR OPERATION             SURVEILLANCE REQUIREMENTS 3.5.G   Automatic    De ressur zat    o          4.5.G   Automatic    De ressurization 3.5.G.3 (Cont'd) vessel pressure shall be reduced to 105 psig or less within 24 hours.
4 CORE AND CONTA NMENT COOLING SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.G Automat c De ressu atio 4.5.G utomatic De ressurizatio 1.
H. Maintenance      of Filled Dischar    e        H. aintenance of   F  lied Dischar e
Four of the six valves of the Automatic Depressurization System shall be OPERABLE:
        ~Pi  e                                              ~Pi e Whenever the core spray systems,                    The  following surveillance LPCI, HPCI, or RCIC are required                    requirements shall be adhered to be OPERABLE, the discharge                      to assure that the discharge piping from the      pump  discharge              piping of the core spray of these    systems to the    last                systems, LPCI, HPCI, and RCIC block valve shall be        filled.                are filled:
1.
The    suction of the RCIC and HPCI                l. Every month and prior to the pumps  shall be aligned to the                         testing of the RHRS (LPCI and condensate storage tank, and                            Containment Spray) and core the pressure suppression chamber                        spray system, the discharge head tank    shall normally be                           piping of,,these systems shall aligned to serve the discharge                          be vented from the high point piping of the RHR and CS pumps.                         and water flow determined.
During each operating cycle the following tests shall be performed on the ADS:
The condensate head tank may be used to serve the RHR and CS                      2. Following any period where discharge piping      if  the PSC head                  the LPCI or core spray systems tank is unavailable. The                                have not been required to be pressure indicators on the                              OPERABLE, the discharge piping discharge of the RHR and CS                              of the inoperable'ystem shall pumps shall indicate not less                            be vented from the high point than  listed  below.                                   prior to the return of the system to service.
(1) PRIOR TO STARTUP from a
Pl-75-20        48 psig Pl-75-48        48 psig                      3. Whenever the HPCI or RCIC Pl-74-51        48 psig                            system is lined up to take Pl-74-65        48 psig                            suction from the condensate storage tank, the discharge piping of the HPCI and RCIC shall be vented from the high point of the system and water flow observed on a monthly basis.
COLD CONDITION, or, (2).whenever there is irradiated fuel in the reactor vessel and the reactor vessel pressure is greater than 105 psig, except in the COLD SHUT-DOWN CONDITION or as specified in 3.5.G.2 and 3.5.G.3 below.
: 4. When  the RHRS  and the CSS are required to be OPERABLE, the pressure indicators which monitor the discharge lines shall be monitored daily and the pressure recorded.
a.
BFN                                           3.5/4.5-17 Unit  2
A simulated automatic actuation test shall be performed PRIOR TO STARTUP after each refueling outage.
Manual surveillance of the relief valves is covered in 4.6.D.2.
: 2. If three of the six ADS valves are known to be incapable of automatic operation, the reactor may remain in operation for a period not to exceed 7.days, provided the HPCI system is OPERABLE. (Note that the pressure relief function of these valves is assured by Section 3.6.D of these specifications and that this specification only applies to the ADS function.) If more than three of the six ADS valves are known to be incapable of automatic operation, an immediate orderly shutdown shall be initiated, with the reactor in a HOT SHUTDOWN CONDITION in 6 hours, and in a COLD SHUTDOWN CONDITION in the following 18 hours.
2.
No additional surveillances are required.
: 3. If Specifications 3.5.G.l and 3.5.G.2 cannot be met, an orderly shutdown will be initiated and the reactor BFN Unit 2 3.5/4.5-16


4    PRIMAR    S S      OUNDAR LIMITING CONDITIONS     FOR OPERATION                 SURVEILLANCE REQUIREMENTS 3.6.C. Coolant Leaka    e                            4.6.C. Coolant Leaka  e lo  a~   Any time    irradiated                      1. Reactor coolant fuel is in the                                   system leakage shall
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.G Automatic De ressur zat o
                ~
4.5.G Automatic De ressurization 3.5.G.3 (Cont'd) vessel pressure shall be reduced to 105 psig or less within 24 hours.
reactor vessel and                               be checked by the reactor coolant                                  sump and air sampling temperature is above                            system and recorded 212'F, reactor coolant                          at least once per leakage into the                                 4 hours.
H.
primary containment from unidentified sources shall not exceed 5 gpm. In addition, the total reactor coolant system leakage into the primary containment shall not exceed 25      gpm.
Maintenance of Filled Dischar e
: b. Anytime the reactor      is in RUN  mode, reactor coolant leakage into the primary containment from unidentified sources shall not increase by more than 2 gpm averaged over any 24-hour period in which the reactor is in the RUN mode except as defined in 3.6.C.l.c below.
~Pi e
c~    During. the first 24 hours in the RUN mode following STARTUP, an    increase in reactor coolant leakage into the primary      "
H.
containment of >2      gpm is acceptable    as long as the requirements of 3.6.C.l.a are met.
aintenance of F lied Dischar e
BFN                                           3.6/4.6-9 AMENDMEHT tl0. I~ 3 Unit  2
~Pi e
Whenever the core spray systems, LPCI, HPCI, or RCIC are required to be OPERABLE, the discharge piping from the pump discharge of these systems to the last block valve shall be filled.
The following surveillance requirements shall be adhered to assure that the discharge piping of the core spray
: systems, LPCI, HPCI, and RCIC are filled:
The suction of the RCIC and HPCI pumps shall be aligned to the condensate storage
: tank, and the pressure suppression chamber head tank shall normally be aligned to serve the discharge piping of the RHR and CS pumps.
The condensate head tank may be used to serve the RHR and CS discharge piping if the PSC head tank is unavailable.
The pressure indicators on the discharge of the RHR and CS pumps shall indicate not less than listed below.
l.
Every month and prior to the testing of the RHRS (LPCI and Containment Spray) and core spray system, the discharge piping of,,these systems shall be vented from the high point and water flow determined.
2.
Following any period where the LPCI or core spray systems have not been required to be OPERABLE, the discharge piping of the inoperable'ystem shall be vented from the high point prior to the return of the system to service.
Pl-75-20 Pl-75-48 Pl-74-51 Pl-74-65 48 psig 48 psig 48 psig 48 psig 3.
Whenever the HPCI or RCIC system is lined up to take suction from the condensate storage
: tank, the discharge piping of the HPCI and RCIC shall be vented from the high point of the system and water flow observed on a monthly basis.
4.
When the RHRS and the CSS are required to be OPERABLE, the pressure indicators which monitor the discharge lines shall be monitored daily and the pressure recorded.
BFN Unit 2 3.5/4.5-17


4   PRIMARY SYSTE      OUNDARY LIMITING CONDITIONS   FOR OPERATION               SURVEILLANCE REQUIREMENTS 4.6.C   Coo ant Leaka e
4 PRIMAR S
: 2. Both the sump and air sampling                2. With the air sampling systems shall be OPERABLE                        system inoperable, grab during  REACTOR POWER OPERATION.                 samples shall be obtained From and after the date that                      and analyzed at least one of these systems is made                      once every 24 hours.
S OUNDAR LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.6.C.
or found to be inoperable for any reason,   REACTOR POWER OPERATION    is permissible only during the succeeding    24 hours for the sump system or 72 hours for the air sampling system.
Coolant Leaka e
The air sampling system may be removed from service for a period of    4 hours for calibration, function testing, and maintenance    without providing    a temporary monitor.
4.6.C.
: 3. If the   condition in   1 or  2 above cannot be met, an orderly shutdown shall be initiated and the reactor shall be placed in the COLD SHUTDOWN CONDITION within  24  hours.
Coolant Leaka e
D. Re  ef Valves
lo a ~
: 1. When more    than one  relief  valve        l. Approximately one-half of is known to be failed, an                        all relief valves shall orderly shutdown shall be                        be bench-checked or initiated and the reactor                        replaced with a depressurized to less than 105                    bench-checked valve psig within 24 hours. The                        each operating cycle.
Any time irradiated fuel is in the
relief valves are not required                    All 13 valves will have to be OPERABLE in the COLD                      been checked or replaced SHUTDOWN    CONDITION.                            upon the completion of every second cycle.
~ reactor vessel and reactor coolant temperature is above 212'F, reactor coolant leakage into the primary containment from unidentified sources shall not exceed 5 gpm.
: 2. In accordance with Specification 1.0.MM, each relief valve shall be manually opened  until thermocouples and acoustic monitors downstream of the valve indicate steam is flowing from the valve.
In addition, the total reactor coolant system leakage into the primary containment shall not exceed 25 gpm.
BFN                                     3.6/4.6-10 Unit  2
1.
Reactor coolant system leakage shall be checked by the sump and air sampling system and recorded at least once per 4 hours.
b.
Anytime the reactor is in RUN mode, reactor coolant leakage into the primary containment from unidentified sources shall not increase by more than 2 gpm averaged over any 24-hour period in which the reactor is in the RUN mode except as defined in 3.6.C.l.c below.
c ~
During. the first 24 hours in the RUN mode following
: STARTUP, an increase in reactor coolant leakage into the primary containment of >2 gpm is acceptable as long as the requirements of 3.6.C.l.a are met.
BFN Unit 2 3.6/4.6-9 AMENDMEHTtl0. I~ 3


s.s/e.s   ~ssEs       0 3.6.B/4.6.C (Cont'd) five gpm, as   specified in 3.6.C, the experimental and analytical data suggest a reasonable margin of safety that such leakage magnitude would not result from a crack approaching the critical size for rapid propagation. Leakage less than the magnitude specified can be detected reasonably in a matter of a few hours utilizing the available leakage detection schemes, and     if the origin cannot be determined in a reasonably short time, the unit should be shut down to allow further investigation   and corrective action.
4 PRIMARY SYSTE OUNDARY LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 4.6.C Coo ant Leaka e
The 2 gpm   limit for coolant   leakage rate increases over any 24-hour period   is a limit specified   by the NRC (Reference 2). This limit applies only during the RUN     mode to avoid being penalized for the expected coolant leakage increase during pressurization.
2.
The   total leakage rate consists of all leakage, identified and unidentified, which flows to the drywell floor drain and equipment drain sumps.
Both the sump and air sampling systems shall be OPERABLE during REACTOR POWER OPERATION.
The capacity of the drywell floor sump pump is 50 gpm and the capacity of the drywell equipment sump pump is also 50 gpm. Removal of 25 gpm from either of these sumps can be accomplished with considerable margin.
From and after the date that one of these systems is made or found to be inoperable for any reason, REACTOR POWER OPERATION is permissible only during the succeeding 24 hours for the sump system or 72 hours for the air sampling system.
2.
With the air sampling system inoperable, grab samples shall be obtained and analyzed at least once every 24 hours.
The air sampling system may be removed from service for a period of 4 hours for calibration, function testing, and maintenance without providing a temporary monitor.
: 3. If the condition in 1 or 2 above cannot be met, an orderly shutdown shall be initiated and the reactor shall be placed in the COLD SHUTDOWN CONDITION within 24 hours.
D.
Re ef Valves 1.
When more than one relief valve is known to be failed, an orderly shutdown shall be initiated and the reactor depressurized to less than 105 psig within 24 hours.
The relief valves are not required to be OPERABLE in the COLD SHUTDOWN CONDITION.
l.
Approximately one-half of all relief valves shall be bench-checked or replaced with a bench-checked valve each operating cycle.
All 13 valves will have been checked or replaced upon the completion of every second cycle.
2.
In accordance with Specification 1.0.MM, each relief valve shall be manually opened until thermocouples and acoustic monitors downstream of the valve indicate steam is flowing from the valve.
BFN Unit 2 3.6/4.6-10
 
s.s/e.s
~ssEs 0
3.6.B/4.6.C (Cont'd) five gpm, as specified in 3.6.C, the experimental and analytical data suggest a reasonable margin of safety that such leakage magnitude would not result from a crack approaching the critical size for rapid propagation.
Leakage less than the magnitude specified can be detected reasonably in a matter of a few hours utilizing the available leakage detection
: schemes, and if the origin cannot be determined in a reasonably short time, the unit should be shut down to allow further investigation and corrective action.
The 2 gpm limit for coolant leakage rate increases over any 24-hour period is a limit specified by the NRC (Reference 2).
This limit applies only during the RUN mode to avoid being penalized for the expected coolant leakage increase during pressurization.
The total leakage rate consists of all leakage, identified and unidentified, which flows to the drywell floor drain and equipment drain sumps.
The capacity of the drywell floor sump pump is 50 gpm and the capacity of the drywell equipment sump pump is also 50 gpm.
Removal of 25 gpm from either of these sumps can be accomplished with considerable margin.
REFERENCE
REFERENCE
: 1. Nuclear System Leakage Rate Limits (BFNP FSAR Subsection 4.10)
: 1. Nuclear System Leakage Rate Limits (BFNP FSAR Subsection 4.10)
: 2. Safety Evaluation Report (SER) on IE Bulletin 82-03 3.6.D/4.6.D     Relief Valves To meet the   safety basis, 13 relief valves have been installed on the unit with   a total capacity of 84.1 percent of nuclear boiler rated steam flow. The analysis of the worst overpressure transient, (3-second closure of all main steam line isolation valves) neglecting the direct scram (valve position scram) results in a maximum vessel pressure which, if a neutron flux scram is assumed considering 12 valves OPERABLE, results in adequate margin to the code allowable overpressure limit of 1,375 psig.
: 2. Safety Evaluation Report (SER) on IE Bulletin 82-03 3.6.D/4.6.D Relief Valves To meet the safety basis, 13 relief valves have been installed on the unit with a total capacity of 84.1 percent of nuclear boiler rated steam flow.
To meet   operational design, the analysis of   the plant isolation transient (generator load reject with bypass     valve failure to open) shows that 12 of the 13 relief valves limit     peak system pressure to a value which is well below the allowed vessel     overpressure of 1,375 psig.
The analysis of the worst overpressure transient, (3-second closure of all main steam line isolation valves) neglecting the direct scram (valve position scram) results in a maximum vessel pressure
Experience in relief valve operation shows that a testing of 50 percent of the valves per year is adequate to detect failures or deteriorations. The relief valves are benchtested every second operating cycle to ensure that their setpoints are within the + 1 percent tolerance. The relief valves are tested in place in accordance with Specification 1.0.MM to establish that they will open and pass steam.
: which, if a neutron flux scram is assumed considering 12 valves
BFN                                         3.6/4.6-30 AMENOMENT 50. IP g Unit  2
: OPERABLE, results in adequate margin to the code allowable overpressure limit of 1,375 psig.
To meet operational design, the analysis of the plant isolation transient (generator load reject with bypass valve failure to open) shows that 12 of the 13 relief valves limit peak system pressure to a
value which is well below the allowed vessel overpressure of 1,375 psig.
Experience in relief valve operation shows that a testing of 50 percent of the valves per year is adequate to detect failures or deteriorations.
The relief valves are benchtested every second operating cycle to ensure that their setpoints are within the
+
1 percent tolerance.
The relief valves are tested in place in accordance with Specification 1.0.MM to establish that they will open and pass steam.
BFN Unit 2 3.6/4.6-30 AMENOMENT50. I P g


3.6/4.6   BASES 3.6.D/4.6.D (Cont'd)
3.6/4.6 BASES 3.6.D/4.6.D (Cont'd)
The requirements established above apply when the nuclear system can be pressurized above ambient conditions. These requirements are applicable at nuclear system pressures below normal operating pressures because abnormal operational transients could possibly start at these conditions such that eventual overpressure relief would be needed.         However, these transients are much less severe, in terms of pressure, than those starting at rated conditions. The valves need not be functional when the vessel head is removed, since the nuclear system cannot be pressurized.
The requirements established above apply when the nuclear system can be pressurized above ambient conditions.
The   relief valves are not required to be OPERABLE in the COLD SHUTDOWN CONDITION.     Overpressure protection is provided during hydrostatic tests by two of the relief valves whose relief setting has been established in conformance with ASME Section ZI code requirements.         The capacity of one relief valve exceeds the charging capacity of the pressurization source used during hydrostatic testing. Two relief valves are used to provide redundancy.
These requirements are applicable at nuclear system pressures below normal operating pressures because abnormal operational transients could possibly start at these conditions such that eventual overpressure relief would be needed.
    ~REPERE RES
: However, these transients are much less severe, in terms of pressure, than those starting at rated conditions.
: 1. Nuclear System Pressure Relief System     (BFNP FSAR Subsection 4.4)
The valves need not be functional when the vessel head is removed, since the nuclear system cannot be pressurized.
: 2. Amendment 22   in response to AEC Question 4.2 of December 6, 1971.
The relief valves are not required to be OPERABLE in the COLD SHUTDOWN CONDITION.
: 3.   "Protection Against Overpressure"     (ASME Boiler and Pressure   Vessel Code, Section III, Article 9)
Overpressure protection is provided during hydrostatic tests by two of the relief valves whose relief setting has been established in conformance with ASME Section ZI code requirements.
: 4. Browns Ferry Nuclear Plant Design Deficiency Report      Target Rock Safety-Relief Valves, transmitted by J. E. Gilleland to F. E. Kruesi, August 29, 1973
The capacity of one relief valve exceeds the charging capacity of the pressurization source used during hydrostatic testing.
: 5. Generic Reload Fuel Application, Licensing Topical Report, NEDE-24011-P-A and Addenda 3.6.E/6.6.E   J~et Pum s Failure of   a jet pump nozzle assembly holddown mechanism, nozzle assembly and/or riser,   would increase the cross-sectional flow area for blowdown following the design basis double-ended line break. Also, failure of the diffuser would eliminate the capability to reflood the core to two-thirds height level following a recirculation line break.         Therefore, if a failure occurred, repairs must be made.
Two relief valves are used to provide redundancy.
The   detection technique is as follows. With the two recirculation pumps balanced   in speed to within g 5 percent, the flow rates in both recirculation loops will be verified by control room monitoring instruments. If the two flow rate values do not differ by more than 10 percent, riser and nozzle assembly integrity has been verified.
~REPERE RES 1.
BFN                                       3.6/4.6-31 Unit  2
Nuclear System Pressure Relief System (BFNP FSAR Subsection 4.4) 2.
Amendment 22 in response to AEC Question 4.2 of December 6, 1971.
3.
"Protection Against Overpressure" (ASME Boiler and Pressure Vessel Code, Section III, Article 9) 4.
Browns Ferry Nuclear Plant Design Deficiency ReportTarget Rock Safety-Relief Valves, transmitted by J. E. Gilleland to F. E. Kruesi, August 29, 1973 5.
Generic Reload Fuel Application, Licensing Topical Report, NEDE-24011-P-A and Addenda 3.6.E/6.6.E J~et Pum s Failure of a jet pump nozzle assembly holddown mechanism, nozzle assembly and/or riser, would increase the cross-sectional flow area for blowdown following the design basis double-ended line break.
Also, failure of the diffuser would eliminate the capability to reflood the core to two-thirds height level following a recirculation line break.
Therefore, if a failure occurred, repairs must be made.
The detection technique is as follows.
With the two recirculation pumps balanced in speed to within g 5 percent, the flow rates in both recirculation loops will be verified by control room monitoring instruments.
If the two flow rate values do not differ by more than 10 percent, riser and nozzle assembly integrity has been verified.
BFN Unit 2 3.6/4.6-31


3.6/4.6   Q~S S 3.6.E/4.6.E (Cont'd)
3.6/4.6 Q~S S
If they   do differ by 10 percent or more, the core flow rate measured by diffuser differential pressure the  jet pump                                    system must be checked against the core flow rate derived from the measured values of loop flow to core flow correlation. If the difference between measured and derived core flow rate is 10 percent or more (with the derived value higher) diffuser measurements will be taken to define the location within the vessel of failed jet pump nozzle (or riser) and the unit shut down for repairs. If the potential blowdown flow area is increased, the system resistance to the recirculation pump is also reduced; hence, the affected drive pump will "run out" to a substantially higher flow rate (approximately 115 percent to 120 percent for a single nozzle failure).
3.6.E/4.6.E (Cont'd)
If   the two loops are balanced in flow at the same pump speed, the resistance characteristics cannot have changed. Any imbalance between drive loop flow rates would be indicated by the plant process instrumentation. In addition, the affected jet pump would provide a leakage path past the core thus reducing the core flow rate. The reverse flow through the inactive jet pump would still be indicated by a positive differential pressure but the net effect would be a slight decrease (3 percent to 6 percent) in the total core flow measured.         This decrease, together with the loop flow increase, would result in a         lack of correlation between measured and derived       core flow rate. Finally, the affected jet pump diffuser     differential   pressure   signal would be reduced because the backflow would be less than the normal forward flow.
If they do differ by 10 percent or more, the core flow rate measured by the jet pump diffuser differential pressure system must be checked against the core flow rate derived from the measured values of loop flow to core flow correlation. If the difference between measured and derived core flow rate is 10 percent or more (with the derived value higher) diffuser measurements will be taken to define the location within the vessel of failed jet pump nozzle (or riser) and the unit shut down for repairs.
A nozzle-riser system failure could also generate the coincident failure of a jet pump diffuser body; however, the converse is not true. The lack of any substantial stress in the jet pump diffuser body makes failure impossible without an initial nozzle-riser system failure.
If the potential blowdown flow area is increased, the system resistance to the recirculation pump is also reduced;
3.6.F/4.6.F   Recirculation   Pum 0 erat on Operation without forced recirculation is permitted for up to 12 hours when the   reactor is not in the RUN mode. And the start of a recirculation pump from the natural circulation condition will not be permitted unless the temperature difference between the loop to be started and the core coolant temperature is less than 75 F. This reduces the positive reactivity insertion to an acceptably low value.
: hence, the affected drive pump will "run out" to a substantially higher flow rate (approximately 115 percent to 120 percent for a single nozzle failure).
Requiring at least one recirculation pump to be operable while in the           RUN mode   provides p'rotection against the potential occurrence of core thermal-hydraulic instabilities at low flow conditions.
If the two loops are balanced in flow at the same pump speed, the resistance characteristics cannot have changed.
Any imbalance between drive loop flow rates would be indicated by the plant process instrumentation.
In addition, the affected jet pump would provide a leakage path past the core thus reducing the core flow rate.
The reverse flow through the inactive jet pump would still be indicated by a positive differential pressure but the net effect would be a slight decrease (3 percent to 6 percent) in the total core flow measured.
This decrease, together with the loop flow increase, would result in a lack of correlation between measured and derived core flow rate.
Finally, the affected jet pump diffuser differential pressure signal would be reduced because the backflow would be less than the normal forward flow.
A nozzle-riser system failure could also generate the coincident failure of a jet pump diffuser body; however, the converse is not true.
The lack of any substantial stress in the jet pump diffuser body makes failure impossible without an initial nozzle-riser system failure.
3.6.F/4.6.F Recirculation Pum 0 erat on Operation without forced recirculation is permitted for up to 12 hours when the reactor is not in the RUN mode.
And the start of a recirculation pump from the natural circulation condition will not be permitted unless the temperature difference between the loop to be started and the core coolant temperature is less than 75 F.
This reduces the positive reactivity insertion to an acceptably low value.
Requiring at least one recirculation pump to be operable while in the RUN mode provides p'rotection against the potential occurrence of core thermal-hydraulic instabilities at low flow conditions.
Requiring the discharge valve of the lower speed loop to remain closed until the speed of the faster pump is below 50% of its rated speed provides assurance when going from one-to-two pump operation that excessive vibration of the jet pump risers will not occur.
Requiring the discharge valve of the lower speed loop to remain closed until the speed of the faster pump is below 50% of its rated speed provides assurance when going from one-to-two pump operation that excessive vibration of the jet pump risers will not occur.
BFN                                     3.6/4.6-32 Unit  2
BFN Unit 2 3.6/4.6-32


3.6/4.6   BASES 3.6.G/4.6.G   St uctu al Inte rit The requirements   for the reactor coolant systems inservice inspection program have been   identified by evaluating the need for a sampling examination of areas of high stress and highest probability of failure in the system and the need to meet as closely as possible the requirements of Section XI, of the   ASME Boiler and Pressure Vessel Code.
3.6/4.6 BASES 3.6.G/4.6.G St uctu al Inte rit The requirements for the reactor coolant systems inservice inspection program have been identified by evaluating the need for a sampling examination of areas of high stress and highest probability of failure in the system and the need to meet as closely as possible the requirements of Section XI, of the ASME Boiler and Pressure Vessel Code.
The program   reflects the built-in limitations of     access to the reactor coolant systems.
The program reflects the built-in limitations of access to the reactor coolant systems.
It is intended that the required examinations and inspection be completed during each 10-year interval. The periodic examinations are to be done during refueling outages or other extended plant shutdown periods.
It is intended that the required examinations and inspection be completed during each 10-year interval.
Only proven nondestructive     testing techniques   will be used.
The periodic examinations are to be done during refueling outages or other extended plant shutdown periods.
More frequent inspections shall be performed on certain circumferential pipe welds as listed in Section 4.6.G.4 to provide additional protection against pipe whip. These welds were selected in respect to their distance from hangers or supports wherein a failure of the weld would permit the unsupported segments of pipe to strike the drywell wall or nearby auxiliary systems or control systems.         Selection was based on judgment from actual plant observation of hanger and support locations and review of drawings. Inspection of all these welds during each 10-year inspection interval will result in three additional examinations above the requirements of Section XI of ASME Code.
Only proven nondestructive testing techniques will be used.
An augmented   inservice surveillance program is required to determine whether any stress corrosion has occurred in any stainless steel piping, stainless components, and highly-stressed alloy steel such as hanger springs, as a result of environmental conditions associated with the March 22, 1975 fire.
More frequent inspections shall be performed on certain circumferential pipe welds as listed in Section 4.6.G.4 to provide additional protection against pipe whip.
REFERENCES
These welds were selected in respect to their distance from hangers or supports wherein a failure of the weld would permit the unsupported segments of pipe to strike the drywell wall or nearby auxiliary systems or control systems.
: 1. Inservice Inspection   and Testing   (BFNP FSAR Subsection 4.12)
Selection was based on judgment from actual plant observation of hanger and support locations and review of drawings.
: 2. Inservice Inspection of Nuclear Reactor Coolant Systems, Section XI, ASME Boiler and Pressure   Vessel Code
Inspection of all these welds during each 10-year inspection interval will result in three additional examinations above the requirements of Section XI of ASME Code.
: 3. ASME Boiler and Pressure   Vessel Code, Section   III (1968 Edition)
An augmented inservice surveillance program is required to determine whether any stress corrosion has occurred in any stainless steel piping, stainless components, and highly-stressed alloy steel such as hanger
: 4. American Society   for Nondestructive Testing     No. SNT-TC-1A (1968 Edition)
: springs, as a result of environmental conditions associated with the March 22, 1975 fire.
: 5. Mechanical Maintenance Instruction 46 (Mechanical Equipment, Concrete, and Structural Steel Cleaning Procedure for Residue From Plant Fire Units 1 and 2)
REFERENCES 1.
: 6. Mechanical Maintenance Instruction 53 (Evaluation of Corrosion       Damage of Piping Components   Which Were Exposed to Residue From March 22, 1975 Fire)
Inservice Inspection and Testing (BFNP FSAR Subsection 4.12) 2.
: 7. Plant Safety Analysis   (BFNP FSAR   Subsection 4.12)
Inservice Inspection of Nuclear Reactor Coolant Systems, Section XI, ASME Boiler and Pressure Vessel Code 3.
BFN                                     3.6/4.6-33 Unit 2
ASME Boiler and Pressure Vessel Code, Section III (1968 Edition) 4.
American Society for Nondestructive Testing No. SNT-TC-1A (1968 Edition) 5.
Mechanical Maintenance Instruction 46 (Mechanical Equipment,
: Concrete, and Structural Steel Cleaning Procedure for Residue From Plant Fire Units 1 and 2) 6.
Mechanical Maintenance Instruction 53 (Evaluation of Corrosion Damage of Piping Components Which Were Exposed to Residue From March 22, 1975 Fire) 7.
Plant Safety Analysis (BFNP FSAR Subsection 4.12)
BFN Unit 2 3.6/4.6-33


UNIT 3 EFFECTIVE PAGE LIST REMOVE                         INSERT 3.2/4.2-14                     3.2/4.2-14 3.2/4.2-15                    3.2/4.2-15 3.2/4.2-22                    3.2/4.2-22*
UNIT 3 EFFECTIVE PAGE LIST REMOVE INSERT 3.2/4.2-14 3.2/4.2-15 3.2/4.2-22 3.2/4.2-23 3.5/4.5-7 3.5/4.5-8 3.5/4.5-12 3.5/4.5-13 3.5/4.5-14 3.5/4.5-15 3.5/4.5-16 3.5/4.5-17 3.6/4.6-9 3.6/4.6-10 3.6/4.6-30 3.6/4.6-31 3.6/4.6-32 3.6/4.6-33 3.2/4.2-14 3.2/4.2-15 3.2/4.2-22*
3.2/4.2-23                    3.2/4.2-23 3.5/4.5-7                      3.5/4.5-7 3.5/4.5-8                      3.5/4.5-8*
3.2/4.2-23 3.5/4.5-7 3.5/4.5-8*
3.5/4.5-12                    3.5/4.5-12*
3.5/4.5-12*
3.5/4.5-13                    3.5/4.5-13 3.5/4.5-14                    3.5/4.5-14 3.5/4.5-15                     3.5/4.5-15*
3.5/4.5-13 3.5/4.5-14 3.5/4.5-15*
3.5/4.5-16                    3.5/4.5-16 3.5/4.5-17                    3.5/4.5-17*
3.5/4.5-16 3.5/4.5-17*
3.6/4.6-9                      3.6/4.6-9*
3.6/4.6-9*
3.6/4.6-10                    3.6/4.6-10 3.6/4.6-30                    3.6/4.6-30*
3.6/4.6-10 3.6/4.6-30*
3.6/4.6-31                    3.6/4.6-31 3.6/4.6-32                    3.6/4.6-32*
3.6/4.6-31 3.6/4.6-32*
3.6/4.6-33                    3.6/4.6-33*
3.6/4.6-33*
+Denotes overleaf or spillover page.
+Denotes overleaf or spillover page.


TABLE 3.2.8 INSTRUMENTATION THAT INITIATES OR CONTROLS THE CORE AND CONTAINHENT COOLING SYSTEHS Hinimum No.
TABLE 3.2.8 INSTRUMENTATION THAT INITIATES OR CONTROLS THE CORE AND CONTAINHENT COOLING SYSTEHS Hinimum No.
Operable Per
Operable Per
~lri Sr 1             Func ion               Tri Level   ettin         Action                      Remarks Instrument Channel-          > 470" above vessel zero.                    1. Below  trip setting initiates Reactor Low Water Level                                                          HPCI .
~lri Sr 1 Func ion Instrument Channel-Reactor Low Water Level Instrument Channel Reactor Low Mater Level Tri Level ettin
Instrument Channel          > 470" above vessel zero.                     1. Multiplier relays initiate Reactor Low Mater Level                                                          RCIC.
> 470" above vessel zero.
Instrument Channel-           > 378" above vessel zero. A               1. Below   trip setting initiates Reactor Low Water Level                                                          CSS.
> 470" above vessel zero.
(LIS-3-58A-D, SW01)
Action Remarks 1.
Below trip setting initiates HPCI 1.
Multiplier relays initiate RCIC.
Instrument Channel-Reactor Low Water Level (LIS-3-58A-D, SW01)
> 378" above vessel zero.
A 1.
Below trip setting initiates CSS.
Hultiplier relays initiate LPCI.
Hultiplier relays initiate LPCI.
: 2. Hultiplier relay from CSS initiates accident signal (15).
2.
2(16)   Instrument Channel-           > 378" above vessel zero. A               1. Below trip settings, in Reactor Low Mater Level                                                         conjunction with drywell (LIS-3-58A-D, SW42)                                                             high pressure, low water level permissive, 120 sec.
Hultiplier relay from CSS initiates accident signal (15).
2(16)
Instrument Channel-
> 378" above vessel zero.
A Reactor Low Mater Level (LIS-3-58A-D, SW42) 1.
Below trip settings, in conjunction with drywell high pressure, low water level permissive, 120 sec.
delay timer and CSS or RHR pump running, initiates ADS.
delay timer and CSS or RHR pump running, initiates ADS.
1(16) Instrument Channel           > 544" above vessel  zero. A              1. Below  trip setting permissive Reactor Low Mater Level                                                         for initiating signals    on ADS.
1(16)
Permissive (LIS-3-184 4 185, SM41)
Instrument Channel Reactor Low Mater Level Permissive (LIS-3-184 4
Instrument Channel           > 312 5/16" above vessel zero. A               1. Below  trip setting prevents Reactor Low Water Level      (2/3 core height)                                 inadvertent operation of (LITS-3-52 and 62, SW41)                                                        containment spray during accident condition.
: 185, SM41)
Instrument Channel Reactor Low Water Level (LITS-3-52 and 62, SW41)
> 544" above vessel zero.
A
> 312 5/16" above vessel zero.
A (2/3 core height) 1.
Below trip setting permissive for initiating signals on ADS.
1.
Below trip setting prevents inadvertent operation of containment spray during accident condition.


TABLE 3.2.B (Continued)
TABLE 3.2.B (Continued)
Hinimum No.
Hinimum No.
Operable Per
Operable Per
~Tri >~~1                 Func i n       Tri  L vel Settin        ~Ac 'I h          Remarks 2(18)     Instrument Channel-   1< p<2.5 psig                       1. Below trip setting prevents Drywell High Pressure                                        inadvertent operation of (PS-64-58 E-H)                                                containment spray during accident conditions.
~Tri >~~1 2(18) 2(18) 2(18) 2(16) (18)
2(18)      Instrument Channel-  <  2.5 psig                          l. Above trip setting in con-Orywell High Pressure                                        junction with low reactor (PS-64-58 A-D, SW&#xb9;2)                                          pressure initiates CSS.
Func i n Instrument Channel-Drywell High Pressure (PS-64-58 E-H)
Instrument Channel-Orywell High Pressure (PS-64-58 A-D, SW&#xb9;2)
Instrument Channel-Drywell High Pressure (PS-64-58A-O, SW&#xb9;1)
Instrument Channel-Orywell High Pressure (PS-64-57A-D)
Tri L vel Settin 1< p<2.5 psig
< 2.5 psig
< 2.5 psig
< 2.5 psig
~Ac
'I h
Remarks 1.
Below trip setting prevents inadvertent operation of containment spray during accident conditions.
l.
Above trip setting in con-junction with low reactor pressure initiates CSS.
Hultiplier relays initiate HPCI.
Hultiplier relays initiate HPCI.
: 2. Hultiplier relay from CSS initiates accident signal.   (15) 2(18)      Instrument Channel-  <  2.5 psig                          l. Above trip setting in Drywell High Pressure                                        conjunction with low (PS-64-58A-O, SW&#xb9;1)                                          reactor pressure initiates LPCI.
2.
2(16) (18) Instrument Channel-  <  2.5 psig                          l. Above trip setting, in Orywell High Pressure                                        conjunction with low reactor (PS-64-57A-D)                                                water level, drywell high pressure, 120 sec. delay timer and CSS or RHR pump running, initiates ADS.
Hultiplier relay from CSS initiates accident signal.
: 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.
(15) l.
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.
Above trip setting in conjunction with low reactor pressure initiates LPCI.
l.
Above trip setting, in conjunction with low reactor water level, drywell high
: pressure, 120 sec.
delay timer and CSS or RHR pump running, initiates ADS.
 
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:
Action:
A. Repair in 24 hours.     If the   function is not   OPERABLE in 24 hours, take action B.
A.
B. Declare the system or component inoperable.
Repair in 24 hours. If the function is not OPERABLE in 24 hours, take action B.
C. Immediately take action     B until power is verified   on the   trip system.
B.
D. No action required; indicators are considered redundant.
Declare the system or component inoperable.
: 2. In only   one trip system.
C.
: 3. Not considered   in a trip system.
Immediately take action B until power is verified on the trip system.
: 4. Requires one channel from each physical location (there are           4 locations) in the steam line space.
D.
: 5. With diesel power, each     RHRS pump   is scheduled to start immediately       and each CSS   pump is sequenced   to start about 7 seconds later.
No action required; indicators are considered redundant.
: 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 RHRS pumps would be operating.
2.
: 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 line break.
In only one trip system.
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.
3.
: 8. Note 1 does not apply to       this item.
Not considered in a trip system.
: 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.
4.
BFN                                             3.2/4.2-22 Unit 3
Requires one channel from each physical location (there are 4 locations) in the steam line space.
5.
With diesel power, each RHRS 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 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 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 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 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 3 3.2/4.2-22


NOTES FOR TABLE     2 B     ontinued)
NOTES FOR TABLE 2
: 10. Only one   trip system for each cooler fan.
B ontinued) 10.
: 11. In only two of the four 4160-V shutdown boards.       See note 13.
Only one trip system for each cooler fan.
: 12. In only one of the four 4160-V shutdown boards.       See note 13.
11.
: 13. An emergency 4160-V shutdown board       is considered a trip system.
In only two of the four 4160-V shutdown boards.
: 14. RHRSW pump   would be inoperable.     Refer to Section 4.5.C for the requirements of a     RHRSW pump   being inoperable.
See note 13.
: 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.
12.
: 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.
In only one of the four 4160-V shutdown boards.
: 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 hours, an orderly power reduction shall be initiated and reactor power shall be less than 30 percent within four hours.
See note 13.
: 18. Not required to be     OPERABLE   in the COLD SHUTDOWN CONDITION.
13.
BFN                                       3.2/4.2-23 Unit 3
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 hours, 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 3 3.2/4.2-23


4   CORE A D CONTA         COOLING SYSTEMS LIMITING CONDITIONS   FOR OPERATION                         SURVEILLANCE REQUIREMENTS 3.5.B   Residual Heat Removal   S ste                     4.5.B   Residua     eat Removal     S ste
4 CORE A
        ~RHRS   (LPCI and Containment                               ~RHRS    (LPCI and Containment Cooling)                                                    Cooling)
D CONTA COOLING SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.B Residual Heat Removal S ste
: 8. If Specifications   3.5.B.l                           8. No  additional surveillance through 3.5.B.7 are not met,                                 required.
~RHRS (LPCI and Containment Cooling) 4.5.B Residua eat Removal S ste
an orderly shutdown shall                 be initiated and the reactor shall be placed in the COLD SHUTDOWN CONDITION within 24 hours.
~RHRS (LPCI and Containment Cooling)
: 9. When  the reactor vessel                                9. When   the reactor vessel pressure is atmospheric and                                 pressure is atmospheric, irradiated fuel is in the                                   the RHR pumps and valves reactor vessel, at least one                                 that are required to be RHR loop with two pumps or two                             OPERABLE    shall  be loops with one pump per loop                                 demonstrated to be shall be OPERABLE. The                                   OPERABLE per diesel generators pumps'ssociated Specification     1.0.MM.
: 8. If Specifications 3.5.B.l through 3.5.B.7 are not met, an orderly shutdown shall be initiated and the reactor shall be placed in the COLD SHUTDOWN CONDITION within 24 hours.
must also be OPERABLE.
8.
: 10. If the conditions of                                   10. No  additional surveillance Specification 3.5.A.5 are met,                               required.
No additional surveillance required.
LPCI and containment   cooling are not required.
9.
ll. When there is irradiated fuel                         11. The  B  and  D RHR  pumps on in the reactor and the reactor                               unit  2  which supply is not in the COLD SHUTDOWN                                 cross-connect capability CONDITION, 2 RHR pumps and                                   shall be demonstrated to associated heat exchangers and                               be OPERABLE per valves on an adjacent unit                                   Specification      1.0.MM when must be OPERABLE and capable                                 the cross-connect of supplying cross-connect                                   capability is required.
When the reactor vessel pressure is atmospheric and irradiated fuel is in the reactor vessel, at least one RHR loop with two pumps or two loops with one pump per loop shall be OPERABLE.
capability except as specified in Specification 3.5.B.12 below. (Note:
The pumps'ssociated diesel generators must also be OPERABLE.
Because cross-connect capability is not a short-term requirement, a component is not considered inoperable if cross-connect capability can be restored to service within 5 hours.)
9.
BFN                                       3.5/4.5-7 Unit  3
When the reactor vessel pressure is atmospheric, the RHR pumps and valves that are required to be OPERABLE shall be demonstrated to be OPERABLE per Specification 1.0.MM.
: 10. If the conditions of Specification 3.5.A.5 are met, LPCI and containment cooling are not required.
10.
No additional surveillance required.
ll.
When there is irradiated fuel in the reactor and the reactor is not in the COLD SHUTDOWN CONDITION, 2 RHR pumps and associated heat exchangers and valves on an adjacent unit must be OPERABLE and capable of supplying cross-connect capability except as specified in Specification 3.5.B.12 below.
(Note:
Because cross-connect capability is not a short-term requirement, a
component is not considered inoperable if cross-connect capability can be restored to service within 5 hours.)
11.
The B and D RHR pumps on unit 2 which supply cross-connect capability shall be demonstrated to be OPERABLE per Specification 1.0.MM when the cross-connect capability is required.
BFN Unit 3 3.5/4.5-7


4     CORE A   CO           COOLING S S EMS LIMITING CONDITIONS     FOR OPERATION               SURVEILLANCE REQUIREMENTS 3.5.B   Res dua Heat Remova S stem                 4.5.B Res dual Heat   Remova S ste
4 CORE A
        ~ggSS. (LPCI and Containment                      iRRHRd   (LPCZ and Containment Cooling)                                         Cooling)
CO COOLING S S
: 12. If one RHR pump or associated             12. No  additional surveillance heat exchanger located                             required.
EMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.B Res dua Heat Remova S stem
on the unit cross-connection in unit 2 is inoperable for any reason   (including valve inoperability, pipe break, etc.), the reactor may remain in operation for a period not to exceed 30 days provided the remaining RHR pump and associated diesel generator are OPERABLE.
~ggSS.
: 13. If RHR cross-connection   flow or         13. No  additional surveillance heat removal capability is lost,                   required.
(LPCI and Containment Cooling) 4.5.B Res dual Heat Remova S ste iRRHRd (LPCZ and Containment Cooling) 12.
the unit may remain in operation for a period not to exceed 10 days unless such capability is restored.
If one RHR pump or associated heat exchanger located on the unit cross-connection in unit 2 is inoperable for any reason (including valve inoperability, pipe break, etc.),
: 14. All recirculation    pump                  14. All recirculation   pump discharge valves shall                            discharge valves shall be OPERABLE PRIOR   TO                           be tested for OPERABILITY STARTUP   (or closed if                           during any period of permitted elsewhere                                COLD SHUTDOWN CONDITION in these specifications).                          exceeding 48 hours, OPERABILITY tests have if not been performed during the preceding 31 days.
the reactor may remain in operation for a period not to exceed 30 days provided the remaining RHR pump and associated diesel generator are OPERABLE.
BFN                                       3.5/4.5-8             AMENDMEHT No. g4 0 Unit  3
12.
No additional surveillance required.
13.
If RHR cross-connection flow or heat removal capability is lost, the unit may remain in operation for a period not to exceed 10 days unless such capability is restored.
13.
No additional surveillance required.
14.
All recirculation pump discharge valves shall be OPERABLE PRIOR TO STARTUP (or closed if permitted elsewhere in these specifications).
14.
All recirculation pump discharge valves shall be tested for OPERABILITY during any period of COLD SHUTDOWN CONDITION exceeding 48 hours, if OPERABILITY tests have not been performed during the preceding 31 days.
BFN Unit 3 3.5/4.5-8 AMENDMEHTNo. g4 0


4     CORE AND CO               COOLI G SYSTEMS LIMITING CONDITIONS       FOR OPERATION           SURVEILLANCE REQUIREMENTS 3.5.C   RHR   Service Water and Emer enc           4.5.C RHR   Service Water and Emer enc E  ui ment Coo in Water S ste s                    ui  ment Coo  n Water   S stems EECWS     Co  t  ued                            EECWS      Continued
4 CORE AND CO COOLI G SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.C RHR Service Water and Emer enc E ui ment Coo in Water S ste s
: 4. One of the Bl     or B2 RHRSW               4. No additional surveillance pumps assigned     to the RHR                     is required.
EECWS Co t ued 4.5.C RHR Service Water and Emer enc ui ment Coo n
heat exchanger supplying the standby coolant supply connection may be inoperable for a period not to exceed 30 days provided the OPERABLE pump is aligned to supply the RHR heat exchanger header and the associated diesel generator and essential control valves are OPERABLE.
Water S stems EECWS Continued 4.
: 5. The standby   coolant supply capability may be inoperable for a period not to exceed 10 days.
One of the Bl or B2 RHRSW pumps assigned to the RHR heat exchanger supplying the standby coolant supply connection may be inoperable for a period not to exceed 30 days provided the OPERABLE pump is aligned to supply the RHR heat exchanger header and the associated diesel generator and essential control valves are OPERABLE.
: 6. If Specifications     3.5.C.2 through 3.5.C.5 are not met, an orderly shutdown shall be initiated     and the unit placed in the     COLD SHUTDOWN CONDITION     within 24 hours.
4.
: 7. There shall   be at least 2 RHRSW pumps,     associated with the selected     RHR pumps, aligned for RHR heat exchanger service for each reactor vessel containing irradiated fuel.
No additional surveillance is required.
BFN                                         3.5/4.5-12             A51ENOMEHT NO. y4 0 Unit  3
5.
The standby coolant supply capability may be inoperable for a period not to exceed 10 days.
: 6. If Specifications 3.5.C.2 through 3.5.C.5 are not
: met, an orderly shutdown shall be initiated and the unit placed in the COLD SHUTDOWN CONDITION within 24 hours.
7.
There shall be at least 2
RHRSW pumps, associated with the selected RHR pumps, aligned for RHR heat exchanger service for each reactor vessel containing irradiated fuel.
BFN Unit 3 3.5/4.5-12 A51ENOMEHT NO. y4 0


LIMITING CONDITIONS     FOR OPERATION           SURVEILLANCE RE UIREMENTS 3.5.D   E ui ment Area Coolers                 4.5.D   E ui ment Area Coolers
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE RE UIREMENTS 3.5.D E ui ment Area Coolers 4.5.D E ui ment Area Coolers 1.
: 1. The equipment area    cooler              l. Each equipment area cooler associated with each RHR                         is operated in conjunction pump and the equipment                           with the equipment served area cooler associated                           by that particular cooler; with each set of core                           therefore, the equipment spray pumps (A and C                             area coolers are tested at or B and D) must be                             the same frequency as the OPERABLE at all times                           pumps which they serve.
The equipment area cooler associated with each RHR pump and the equipment area cooler associated with each set of core spray pumps (A and C
when the pump or pumps served by that specific cooler is considered to be OPERABLE.
or B and D) must be OPERABLE at all times when the pump or pumps served by that specific cooler is considered to be OPERABLE.
: 2. When an equipment area cooler is not OPERABLE, the pump(s) served by that cooler must be considered inoperable for Technical Specification purposes.
l.
E. Hi h Pressure     Coolant In ection             E. Hi h P essure   Coo ant S  ste    HPCIS                                      In ection S stem   HPCIS
Each equipment area cooler is operated in conjunction with the equipment served by that particular cooler; therefore, the equipment area coolers are tested at the same frequency as the pumps which they serve.
: 1. The HPCI system   shall be                     1. HPCI Subsystem    testing OPERABLE   whenever there is                         shall be performed as irradiated fuel in the                               follows:
2.
reactor vessel and the reactor vessel pressure                         a. Simulated        Once/18 is greater than 150 psig,                           Automatic        months except   in the COLD SHUTDOWN                       Actuation CONDITION or as specified in                         Test 3.5.E.2. OPERABILITY shall be determined   within                         b. Pump            Per 12 hours after reactor                             OPERA-          Specification steam pressure reaches                               BILITY          1.0.MM 150 psig from a COLD CONDITION, or alternatively                     c. Motor Qper-      Per PRIOR TO STARTUP by using an                         ated Valve      Specification auxiliary   steam supply.                           OPERABILITY     1.0.MM
When an equipment area cooler is not OPERABLE, the pump(s) served by that cooler must be considered inoperable for Technical Specification purposes.
: d. Flow Rate at     Once/3 normal           months reactor vessel operating pressure BFN                                       3.5/4.5-13 Unit  3
E.
Hi h Pressure Coolant In ection S ste HPCIS E.
Hi h P essure Coo ant In ection S stem HPCIS 1.
The HPCI system shall be OPERABLE whenever there is irradiated fuel in the reactor vessel and the reactor vessel pressure is greater than 150 psig, except in the COLD SHUTDOWN CONDITION or as specified in 3.5.E.2.
OPERABILITY shall be determined within 12 hours after reactor steam pressure reaches 150 psig from a COLD CONDITION, or alternatively PRIOR TO STARTUP by using an auxiliary steam supply.
a.
Simulated Automatic Actuation Test Once/18 months b.
Pump OPERA-BILITY Per Specification 1.0.MM c.
Motor Qper-ated Valve OPERABILITY Per Specification 1.0.MM 1.
HPCI Subsystem testing shall be performed as follows:
d.
Flow Rate at Once/3 normal months reactor vessel operating pressure BFN Unit 3 3.5/4.5-13


.5/4   5 CORE A   CO   AI       COOLI G SYSTEMS LIMITING CONDITIONS     FOR OPERATION             SURVEILLANCE REQUIREMENTS 3.5.E     Hi h Pressure   Coo ant n ect o       4.5.E     Hi h Pressure Coolant In 'ectio 4.5.E.l   (Cont'd)
.5/4 5
: e. Flow Rate at   Once/18 150 psig       months The HPCI pump   shall deliver at least 5000   gpm during each flow rate test.
CORE A
: f. Verify that       Once/Month each valve (manual, power-operated, or automatic) in the injection flow-path that is not locked, sealed, or otherwise secured in position, is in its correct* position.
CO AI COOLI G SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.E Hi h Pressure Coo ant n ect o
: 2. If the HPCI system   is                   2. No additional surveillances inoperable, the reactor may                     are required.
4.5.E Hi h Pressure Coolant In 'ectio 4.5.E.l (Cont'd) e.
remain in operation for a period not to exceed 7 days, provided the ADS,. CSS, RHRS (LPCI), and RCICS are OPERABLE.
Flow Rate at Once/18 150 psig months The HPCI pump shall deliver at least 5000 gpm during each flow rate test.
: 3. If Specifications   3.5.E.l
f.
* Except that an automatic or 3.5.E.2 are not met,                           valve capable of automatic an  orderly shutdown shall                        return to its ECCS position be  initiated  and the                            when an ECCS signal is reactor vessel pressure                            present may be in a shall be reduced to 150                            position for another   mode psig or less within 24                            of operation.
Verify that Once/Month each valve (manual, power-
hours.
: operated, or automatic) in the injection flow-path that is not locked, sealed, or otherwise secured in position, is in its correct* position.
F. Reactor Core   Isolation Coolin           F. Reactor Core Isolation Coolin
: 2. If the HPCI system is inoperable, the reactor may remain in operation for a period not to exceed 7 days, provided the ADS,. CSS, RHRS (LPCI), and RCICS are OPERABLE.
: 1. The RCICS   shall be OPERABLE             1. RCIC Subsystem  testing shall whenever there   is irradiated               be performed as  follows:
2.
fuel in the reactor vessel and the   reactor vessel                       a. Simulated Auto- Once/18 pressure is above 150 psig,                       matic Actuation months except in the COLD SHUTDOWN                       Test CONDITION or as specified in 3.5.F.2. OPERABILITY shall BFN                                       3.5/4.5-14 Unit  3
No additional surveillances are required.
: 3. If Specifications 3.5.E.l or 3.5.E.2 are not met, an orderly shutdown shall be initiated and the reactor vessel pressure shall be reduced to 150 psig or less within 24 hours.
Except that an automatic valve capable of automatic return to its ECCS position when an ECCS signal is present may be in a position for another mode of operation.
F.
Reactor Core Isolation Coolin F.
Reactor Core Isolation Coolin 1.
The RCICS shall be OPERABLE whenever there is irradiated fuel in the reactor vessel and the reactor vessel pressure is above 150 psig, except in the COLD SHUTDOWN CONDITION or as specified in 3.5.F.2.
OPERABILITY shall 1.
RCIC Subsystem testing shall be performed as follows:
a.
Simulated Auto-Once/18 matic Actuation months Test BFN Unit 3 3.5/4.5-14


4     CORE   D CO LIMITING CONDITIONS COO FOR OPERATION G S  STEMS        t SURVEILLANCE REQUIREMENTS 3.5.F   eacto Core Iso atio Coolin           4.5.F Reactor Core   Isolat on Coolin 3.5.F.l (Cont'd)                            4.5.F.l (Cont'd) be determined   within 12 hours           b. Pump                Per after reactor steam pressure                 OPERABILITY        Specifi-reaches 150 psig from a COLD                                     cation CONDITION or alternatively                                       1.0.MM PRIOR TO STARTUP by using an auxiliary steam supply.                   c. Motor-Operated     Per Valve              Specifi-OPERABILITY        cation 1.0.MM
4 CORE D
: d. Flow Rate at       Once/3 normal reactor     months vessel operating pressure
CO t
: e. Flow Rate at       Once/18 150  psig          months The RCIC pump   shall deliver at least 600 gpm during each flow test.
COO G
: 2. If the RCICS is inoperable,             f. Verify that        Once/Month'ach the reactor may remain in                             valve operation for a period not                     (manual, power-to exceed 7 days if the                      operated, or automatic) in the HPCIS is OPERABLE during such time.                                     injection flowpath that is not locked,
S STEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.F eacto Core Iso atio Coolin 4.5.F Reactor Core Isolat on Coolin 3.5.F.l (Cont'd) be determined within 12 hours after reactor steam pressure reaches 150 psig from a COLD CONDITION or alternatively PRIOR TO STARTUP by using an auxiliary steam supply.
: 3. If Specifications 3.5.F.l                     sealed, or other-wise secured in or 3.5.F.2 are not met, an orderly shutdown shall be                     position, is in its initiated and the reactor                     correct* position.
4.5.F.l (Cont'd) b.
shall be depressurized to less than 150 psig within               2. No additional surveillances 24 hours.                                  are required.
Pump OPERABILITY
* Except that     an automatic valve capable of automatic return to its normal position when a signal is present may be in a position for another mode of operation.
: c. Motor-Operated Valve OPERABILITY Per Specifi-cation 1.0.MM Per Specifi-cation 1.0.MM d.
BFN                                     3.5/4.5-15               Al"lENDMEi)TNo Unit  3
Flow Rate at Once/3 normal reactor months vessel operating pressure e.
Flow Rate at 150 psig Once/18 months The RCIC pump shall deliver at least 600 gpm during each flow test.
: 2. If the RCICS is inoperable, the reactor may remain in operation for a period not to exceed 7 days if the HPCIS is OPERABLE during such time.
: 3. If Specifications 3.5.F.l or 3.5.F.2 are not met, an orderly shutdown shall be initiated and the reactor shall be depressurized to less than 150 psig within 24 hours.
f.
Verify that Once/Month'ach valve (manual, power-
: operated, or automatic) in the injection flowpath that is not locked,
: sealed, or other-wise secured in position, is in its correct* position.
2.
No additional surveillances are required.
Except that an automatic valve capable of automatic return to its normal position when a signal is present may be in a position for another mode of operation.
BFN Unit 3 3.5/4.5-15 Al"lENDMEi)TNo


.5 4     CORE AND CO     A         COOLING SYS E   S LIMITING CONDITIONS     FOR OPERATION               SURVEILLANCE REQUIREMENTS 3.5.G   Automatic   De   ressur zation             4.5.G   Automatic De ressurization Four of the     six valves of                 1. During each operating the Automatic                                      cycle the following Depressurization System                            tests shall be performed shall  be OPERABLE:                                on the ADS:
.5 4 CORE AND CO A
(1) PRIOR TO STARTUP from                         a. A simulated automatic a COLD CONDlTION, or,                               actuation test shall be performed PRIOR  TO (2) whenever there is                                   STARTUP after each irradiated fuel in the                             refueling outage.
COOLING SYS E
reactor vessel and the                             Manual  surveillance reactor vessel pressure                             of the relief valves is greater than 105 psig,                           is covered in except     in the COLD SHUT-                       4.6.D.2.
S LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.G Automatic De ressur zation 4.5.G Automatic De ressurization Four of the six valves of the Automatic Depressurization System shall be OPERABLE:
DOWN   CONDITION   or as specified in 3.5.G.2       and 3.5.G.3 below.
1.
: 2. If three   .of the six ADS                     2. No additional surveillances valves are known to be                             are required.
During each operating cycle the following tests shall be performed on the ADS:
incapable of automatic operation, the reactor may remain in operation for a period not to exceed 7 days, provided the HPCI system is OPERABLE. (Note that the pressure relief function of these valves is assured by Section 3.6.D of these specifications and that this specification only applies to the ADS function.) If more than three of the six ADS valves are known to be incapable of automatic operation, an immediate orderly shutdown shall be initiated, with the reactor in a HOT SHUTDOWN CONDITION in 6 hours, and in a COLD SHUTDOWN CONDITION in the following 18 hours.
(1) PRIOR TO STARTUP from a
3 ~ If Specifications       3.5.G.1 and 3.5.G.2 cannot be met, an orderly shutdown       will be initiated     and the   reactor BFN                                         3.5/4.5-16 Unit 3
COLD CONDlTION, or, (2) whenever there is irradiated fuel in the reactor vessel and the reactor vessel pressure is greater than 105 psig, except in the COLD SHUT-DOWN CONDITION or as specified in 3.5.G.2 and 3.5.G.3 below.
: 4. CO E      CO            COOLI  G SYS  E S LIMITING CONDITIONS    FOR OPERATION              SURVEILLANCE REQUIREMENTS 3.5.G  Automat c De    ressurization            4.5.G  Automatic    De ressurization S stem    ADS 3.5.G.3 (Cont'd) vessel pressure shall be reduced to 105 psig or less within 24 hours.
a.
H. Maintenance    of Filled Dischar    e          H. Maintenance of    Filled Dischar e
A simulated automatic actuation test shall be performed PRIOR TO STARTUP after each refueling outage.
        ~Pi e                                              ~Pi e Whenever the core spray systems,                  The  following surveillance LPCI, HPCI, or RCIC are required                  requirements shall be adhered to be OPERABLE, the discharge                      to assure that the discharge piping from the    pump  discharge                piping of the core spray of these    systems to the    last                systems, LPCI, HPCI, and RCIC block valve shall be      filled.                  are filled:
Manual surveillance of the relief valves is covered in 4.6.D.2.
The  suction of the RCIC and HPCI                1. Every month and prior to the pumps  shall be aligned to the                          testing of the RHRS (LPCI and condensate storage tank, and                            Containment Spray) and core the pressure suppression chamber                        spray systems, the discharge head tank    shall normally be                          piping of these systems shall aligned to serve the discharge                          be vented from the high point piping of the RHR and CS pumps.                          and water flow determined.
: 2. If three.of the six ADS valves are known to be incapable of automatic operation, the reactor may remain in operation for a period not to exceed 7 days, provided the HPCI system is OPERABLE. (Note that the pressure relief function of these valves is assured by Section 3.6.D of these specifications and that this specification only applies to the ADS function.) If more than three of the six ADS valves are known to be incapable of automatic operation, an immediate orderly shutdown shall be initiated, with the reactor in a HOT SHUTDOWN CONDITION in 6 hours, and in a COLD SHUTDOWN CONDITION in the following 18 hours.
The condensate head tank may be used to serve the RHR and CS                      2. Following any period where the discharge piping    if  the PSC head                    LPCI  or core spray systems tank is unavailable. The                                have  not been required to be pressure indicators on the                              OPERABLE, the discharge piping discharge of the RHR and CS pumps                        of the inoperable system shall shall indicate not less than                            be vented from the high point listed below.                                            prior to the return of the system to service.
2.
Pl-75-20      48  psig Pl-75-48      48  psig                    3. Whenever the HPCI or RCIC Pl-74-51      48  psig                          system is lined up to take P1-74-65      48  psig                          suction from the condensate storage tank, the discharge piping of the HPCI and RCIC shall be vented from the high point of the system and water flow observed on a monthly basis.
No additional surveillances are required.
: 4. When  the RHRS and the CSS are required to be OPERABLE, the pressure indicators which monitor the discharge lines shall be monitored daily and the pressure recorded.
3 ~ If Specifications 3.5.G.1 and 3.5.G.2 cannot be met, an orderly shutdown will be initiated and the reactor BFN Unit 3 3.5/4.5-16
BFN                                        3.5/4.5-17 Unit  3


6 4 6    RIMARY SYST      OUNDARY LIMITING CONDITIONS   FOR OPERATION               SURVEILLANCE REQUIREMENTS 4.6.C  Coola t  Leaka e
4.
: l. a. Any time  irradiated                    1. Reactor coolant fuel is in the                               system leakage shall reactor vessel and                           be checked by the reactor coolant                              sump and air sampling temperature is above                          system and recorded 212'F, reactor coolant                        at least once per leakage into the                             4 hours.
CO E
primary containment from unidentified sources shall not exceed 5 gpm. In addition, the total reactor coolant system leakage into the primary containment shall not   'exceed 25 gpm.
CO COOLI G
: b. Anytime the reactor      is in RUN  mode,   reactor coolant leakage into the primary containment from unidentified sources shall not increase by more than 2 gpm averaged over any 24-hour period in which the reactor is in the RUN mode except as defined in 3.6.C.l.c below.
SYS E
: c. During the   first  24 hours in the RUN    mode  following STARTUP, an    increase in reactor coolant leakage into the primary containment of >2    gpm is acceptable as long as the requirements of 3.6.C.l.a are met.
S LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.G Automat c De ressurization S stem ADS 4.5.G Automatic De ressurization 3.5.G.3 (Cont'd) vessel pressure shall be reduced to 105 psig or less within 24 hours.
BFN                                       3.6/4.6-9 AMEND",ATE/ lT NO. Z0 8 Unit 3
H.
Maintenance of Filled Dischar e
~Pi e
H.
Maintenance of Filled Dischar e
~Pi e
Whenever the core spray systems, LPCI, HPCI, or RCIC are required to be OPERABLE, the discharge piping from the pump discharge of these systems to the last block valve shall be filled.
The following surveillance requirements shall be adhered to assure that the discharge piping of the core spray
: systems, LPCI, HPCI, and RCIC are filled:
The suction of the RCIC and HPCI pumps shall be aligned to the condensate storage
: tank, and the pressure suppression chamber head tank shall normally be aligned to serve the discharge piping of the RHR and CS pumps.
The condensate head tank may be used to serve the RHR and CS discharge piping if the PSC head tank is unavailable.
The pressure indicators on the discharge of the RHR and CS pumps shall indicate not less than listed below.
1.
Every month and prior to the testing of the RHRS (LPCI and Containment Spray) and core spray systems, the discharge piping of these systems shall be vented from the high point and water flow determined.
2.
Following any period where the LPCI or core spray systems have not been required to be OPERABLE, the discharge piping of the inoperable system shall be vented from the high point prior to the return of the system to service.
Pl-75-20 Pl-75-48 Pl-74-51 P1-74-65 48 psig 48 psig 48 psig 48 psig 3.
Whenever the HPCI or RCIC system is lined up to take suction from the condensate storage
: tank, the discharge piping of the HPCI and RCIC shall be vented from the high point of the system and water flow observed on a monthly basis.
4.
When the RHRS and the CSS are required to be OPERABLE, the pressure indicators which monitor the discharge lines shall be monitored daily and the pressure recorded.
BFN Unit 3 3.5/4.5-17


4     RIMARY SYST   . OUNDARY LIMITING CONDITIONS   FOR OPERATION             SURVEILLANCE REQUIREMENTS 3.6.C   Coolant Leaka   e                       4.6.C   Coolant Leaka e
6 4 6
: 2. Both the sump and air sampling              2. With the air sampling systems shall be OPERABLE                       system inoperable, grab during REACTOR POWER OPERATION.                 samples shall be obtained From and after the date that                     and analyzed at least one of these systems is made or                 once every 24 hours.
RIMARY SYST OUNDARY LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 4.6.C Coola t Leaka e
found to be inoperable for any reason,   REACTOR POWER OPERATION is permissible only during the succeeding 24 hours for the sump system or 72 hours for the air sampling system.
l.
The air sampling system may be removed from service for a period of 4 hours for calibration, function testing, and maintenance without providing a temporary monitor.
a.
: 3. If the condition in   1 or 2 above cannot be met, an orderly shutdown shall be initiated and the reactor shall be placed in the COLD SHUTDOWN CONDITION within 24 hours.
Any time irradiated fuel is in the reactor vessel and reactor coolant temperature is above 212'F, reactor coolant leakage into the primary containment from unidentified sources shall not exceed 5 gpm.
: 1. When more   than one relief   valve         l. Approximately one-half of is known to be failed, an                     all relief valves shall orderly shutdown shall be                       be bench-checked or initiated and the reactor                       replaced with a depressurized to less than 105                   bench-checked valve psig within 24 hours. The                       each operating cycle.
In addition, the total reactor coolant system leakage into the primary containment shall not 'exceed 25 gpm.
relief valves are not required                 All 13 valves will have to be OPERABLE in the COLD                       been checked or replaced SHUTDOWN  CONDITION.                          upon the completion of every second cycle.
1.
: 2. In accordance with Specification 1.0.MM, each relief valve shall be manually opened until thermocouples and acoustic monitors downstream of the valve indicate steam is flowing from the valve.
Reactor coolant system leakage shall be checked by the sump and air sampling system and recorded at least once per 4 hours.
BFN                                     3.6/4.6-10 Unit  3
b.
Anytime the reactor is in RUN mode, reactor coolant leakage into the primary containment from unidentified sources shall not increase by more than 2 gpm averaged over any 24-hour period in which the reactor is in the RUN mode except as defined in 3.6.C.l.c below.
c.
During the first 24 hours in the RUN mode following
: STARTUP, an increase in reactor coolant leakage into the primary containment of >2 gpm is acceptable as long as the requirements of 3.6.C.l.a are met.
BFN Unit 3 3.6/4.6-9 AMEND",ATE/lT NO. Z 0 8
 
4 RIMARY SYST OUNDARY LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.6.C Coolant Leaka e
4.6.C Coolant Leaka e
2.
Both the sump and air sampling systems shall be OPERABLE during REACTOR POWER OPERATION.
From and after the date that one of these systems is made or found to be inoperable for any
: reason, REACTOR POWER OPERATION is permissible only during the succeeding 24 hours for the sump system or 72 hours for the air sampling system.
2.
With the air sampling system inoperable, grab samples shall be obtained and analyzed at least once every 24 hours.
The air sampling system may be removed from service for a period of 4 hours for calibration, function
: testing, and maintenance without providing a temporary monitor.
: 3. If the condition in 1 or 2 above cannot be met, an orderly shutdown shall be initiated and the reactor shall be placed in the COLD SHUTDOWN CONDITION within 24 hours.
1.
When more than one relief valve is known to be failed, an orderly shutdown shall be initiated and the reactor depressurized to less than 105 psig within 24 hours.
The relief valves are not required to be OPERABLE in the COLD SHUTDOWN CONDITION.
l.
Approximately one-half of all relief valves shall be bench-checked or replaced with a bench-checked valve each operating cycle.
All 13 valves will have been checked or replaced upon the completion of every second cycle.
2.
In accordance with Specification 1.0.MM, each relief valve shall be manually opened until thermocouples and acoustic monitors downstream of the valve indicate steam is flowing from the valve.
BFN Unit 3 3.6/4.6-10
 
o


o 3.6/4.6   QASES 3.6.C/4.6.C (Cont'd) suggest a reasonable margin of safety that such leakage magnitude would not result from a crack approaching the critical size for rapid propagation. Leakage less than the magnitude specified can be detected reasonably in a matter of a few hours utilizing the available leakage detection schemes, and     if the origin cannot be determined in a reasonably short time, the unit should be shut down to allow further investigation and   corrective action.
3.6/4.6 QASES 3.6.C/4.6.C (Cont'd) suggest a reasonable margin of safety that such leakage magnitude would not result from a crack approaching the critical size for rapid propagation.
The two gpm     limit for coolant leakage rate increase over any 24 hour period is a limit specified by the NRC (Reference 2). This limit applies only duri'ng the RUN mode to avoid being penalized for the expected coolant leakage increase during pressurization.
Leakage less than the magnitude specified can be detected reasonably in a matter of a few hours utilizing the available leakage detection
The   total leakage rate consists of all leakage, identified and unidentified, which flows to the drywell floor drain and equipment drain sumps ~
: schemes, and if the origin cannot be determined in a reasonably short time, the unit should be shut down to allow further investigation and corrective action.
The   capacity of the drywell floor sump pump is 50 gpm and the capacity of the drywell equipment sump pump is also 50 gpm. Removal of 25 gpm from either of these sumps can be accomplished with considerable margin.
The two gpm limit for coolant leakage rate increase over any 24 hour period is a limit specified by the NRC (Reference 2).
References
This limit applies only duri'ng the RUN mode to avoid being penalized for the expected coolant leakage increase during pressurization.
: l. Nuclear System Leakage Rate Limits (BFNP FSAR Subsection 4.10)
The total leakage rate consists of all leakage, identified and unidentified, which flows to the drywell floor drain and equipment drain sumps
: 2. Safety Evaluation Report (SER) on IE Bulletin 82-03 D   4   D Relief Valves To meet   the safety basis,   13 relief valves have been installed on the unit with a total capacity of 83.77 percent of nuclear boiler rated steam flow. The analysis of the worst overpressure transient, (3-second closure of all main steam line isolation valves) neglecting the direct scram (valve position scram) results in a maximum vessel pressure which, if a neutron flux scram is assumed considering 12 valves OPERABLE, results in adequate margin to the code allowable overpressure limit of 1,375 psig.
~
To meet   operational design, the analysis of the plant isolation transient (generator load reject with bypass valve failure to open) shows that 12 of the 13 relief valves limit peak system pressure to a value which is well below the allowed vessel overpressure of 1,375 psig.
The capacity of the drywell floor sump pump is 50 gpm and the capacity of the drywell equipment sump pump is also 50 gpm.
Experience in relief and safety valve operation shows that a testing of 50 percent of the valves per year is adequate to detect failures or deteriorations. The relief and safety valves are benchtested every second operating cycle to ensure that their setpoints are within the g 1 percent tolerance. The relief valves are tested in place in accordance with Specification 1.0.MM to establish that they will open and pass steam.
Removal of 25 gpm from either of these sumps can be accomplished with considerable margin.
BFN                                       3.6/4.6-30           AMEND%!EAT HO. I4 I Unit
References l.
2.
Nuclear System Leakage Rate Limits (BFNP FSAR Subsection 4.10)
Safety Evaluation Report (SER) on IE Bulletin 82-03 D 4 D
Relief Valves To meet the safety basis, 13 relief valves have been installed on the unit with a total capacity of 83.77 percent of nuclear boiler rated steam flow.
The analysis of the worst overpressure transient, (3-second closure of all main steam line isolation valves) neglecting the direct scram (valve position scram) results in a maximum vessel pressure
: which, if a neutron flux scram is assumed considering 12 valves
: OPERABLE, results in adequate margin to the code allowable overpressure limit of 1,375 psig.
To meet operational
: design, the analysis of the plant isolation transient (generator load reject with bypass valve failure to open) shows that 12 of the 13 relief valves limit peak system pressure to a value which is well below the allowed vessel overpressure of 1,375 psig.
Experience in relief and safety valve operation shows that a testing of 50 percent of the valves per year is adequate to detect failures or deteriorations.
The relief and safety valves are benchtested every second operating cycle to ensure that their setpoints are within the g
1 percent tolerance.
The relief valves are tested in place in accordance with Specification 1.0.MM to establish that they will open and pass steam.
BFN Unit 3.6/4.6-30 AMEND%!EAT HO. I4 I


3.6/4.6   BASES 3.6.D/4.6.D (Cont'd)
3.6/4.6 BASES 3.6.D/4.6.D (Cont'd)
The requirements     established above apply when the nuclear system can be pressurized above ambient conditions. These requirements are applicable at nuclear system pressures below normal operating pressures because abnormal operational transients could possibly start at these conditions such that eventual overpressure relief would be needed. However, these transients are much less severe, in terms of pressure, than those starting at rated conditions. The valves need not be functional when the vessel head is removed, since the nuclear system cannot be pressurized.
The requirements established above apply when the nuclear system can be pressurized above ambient conditions.
The relief   valves are not required to be OPERABLE in the COLD SHUTDOWN CONDITION. Overpressure protection is provided during hydrostatic tests by two of the   relief valves whose relief setting has been established in conformance with ASME Section XI code requirements.           The capacity of one relief valve exceeds the charging capacity of       the   pressurization   source used during hydrostatic testing. Two relief       valves   are used to provide redundancy; References
These requirements are applicable at nuclear system pressures below normal operating pressures because abnormal operational transients could possibly start at these conditions such that eventual overpressure relief would be needed.
: 1. Nuclear System Pressure Relief System       (BFNP FSAR   Subsection 4.4)
: However, these transients are much less severe, in terms of pressure, than those starting at rated conditions.
: 2.   "Protection Against Overpressure"     (ASME Boiler   and Pressure   Vessel Code, Section   III, Article 9)
The valves need not be functional when the vessel head is
: 3. Browns   Ferry Nuclear Plant Design Deficiency Report        Target Rock Safety-Relief Valves, transmitted by J.       E. Gilliland   to F. E. Kruesi, August 29, 1973 3.6.E/4.6.E     3~et Pum s Failure of   a jet pump nozzle assembly holddown mechanism, nozzle assembly and/or riser, would increase the cross-sectional flow area for blowdown following the design basis double-ended line break. Also, failure of the diffuser would eliminate the capability to reflood the core to two-thirds height level following a recirculation line break. Therefore, if a failure occurred, repairs must be made.
: removed, since the nuclear system cannot be pressurized.
The detection technique is as follows. With the two recirculation pumps balanced in speed to within g 5 percent, the flow rates in both recirculation loops will be verified by control room monitoring instruments.             If the two flow rate values do not differ by more than 10 percent, riser and             nozzle assembly integrity has been verified.
The relief valves are not required to be OPERABLE in the COLD SHUTDOWN CONDITION.
If they   do differ by 10 percent or more, the core flow rate measured by the jet pump diffuser differential pressure system must be checked against the core flow rate derived from the measured values of loop flow to core flow correlation.     If the difference between measured and derived core flow rate is 10 percent or more (with the derived value higher) diffuser measurements will be taken to define the location within the vessel of failed jet pump nozzle (or riser) and the unit shut down for repairs.         If the potential blowdown flow BFN                                       3.6/4.6-31 Unit  3
Overpressure protection is provided during hydrostatic tests by two of the relief valves whose relief setting has been established in conformance with ASME Section XI code requirements.
The capacity of one relief valve exceeds the charging capacity of the pressurization source used during hydrostatic testing.
Two relief valves are used to provide redundancy; References 1.
Nuclear System Pressure Relief System (BFNP FSAR Subsection 4.4) 2.
"Protection Against Overpressure" (ASME Boiler and Pressure Vessel
: Code, Section III, Article 9) 3.
Browns Ferry Nuclear Plant Design Deficiency ReportTarget Rock Safety-Relief Valves, transmitted by J. E. Gilliland to F. E. Kruesi, August 29, 1973 3.6.E/4.6.E 3~et Pum s Failure of a jet pump nozzle assembly holddown mechanism, nozzle assembly and/or riser, would increase the cross-sectional flow area for blowdown following the design basis double-ended line break.
Also, failure of the diffuser would eliminate the capability to reflood the core to two-thirds height level following a recirculation line break.
Therefore, if a failure
: occurred, repairs must be made.
The detection technique is as follows.
With the two recirculation pumps balanced in speed to within g 5 percent, the flow rates in both recirculation loops will be verified by control room monitoring instruments.
If the two flow rate values do not differ by more than 10 percent, riser and nozzle assembly integrity has been verified.
If they do differ by 10 percent or more, the core flow rate measured by the jet pump diffuser differential pressure system must be checked against the core flow rate derived from the measured values of loop flow to core flow correlation. If the difference between measured and derived core flow rate is 10 percent or more (with the derived value higher) diffuser measurements will be taken to define the location within the vessel of failed jet pump nozzle (or riser) and the unit shut down for repairs.
If the potential blowdown flow BFN Unit 3 3.6/4.6-31


3.6/4.6   BASES 3.6.E/4.6.E (Cont'd) area is increased, the system resistance       to the recirculation pump is also reduced; hence, the affected drive pump       will "run out" to a substantially higher flow rate (approximately 115 percent to 120 percent for a single nozzle failure). If the two loops are balanced in flow at the same pump speed, the resistance characteristics cannot have 'changed. Any imbalance between drive loop flow rates would be indicated by the plant process instrumentation. In addition, the affected jet pump would provide a leakage path past the core thus reducing the core flow rate. The reverse flow through the inactive jet pump would still be indicated by a positive differential pressure but the net effect would be a slight decrease (3 percent to 6 percent) in the total core flow measured. This decrease, together with the loop flow increase, would result in a lack of correlation between measured and derived core flow rate.
3.6/4.6 BASES 3.6.E/4.6.E (Cont'd) area is increased, the system resistance to the recirculation pump is also reduced;
: hence, the affected drive pump will "run out" to a substantially higher flow rate (approximately 115 percent to 120 percent for a single nozzle failure). If the two loops are balanced in flow at the same pump speed, the resistance characteristics cannot have 'changed.
Any imbalance between drive loop flow rates would be indicated by the plant process instrumentation.
In
: addition, the affected jet pump would provide a leakage path past the core thus reducing the core flow rate.
The reverse flow through the inactive jet pump would still be indicated by a positive differential pressure but the net effect would be a slight decrease (3 percent to 6 percent) in the total core flow measured.
This decrease, together with the loop flow increase, would result in a lack of correlation between measured and derived core flow rate.
Finally, the affected jet pump diffuser differential pressure signal would be reduced because the backflow would be less than the normal forward flow.
Finally, the affected jet pump diffuser differential pressure signal would be reduced because the backflow would be less than the normal forward flow.
A nozzle-riser system failure could also generate the coincident failure of         a jet pump diffuser body; however, the converse is not true. The lack of any substantial stress in the jet pump diffuser body makes failure impossible without an initial nozzle-riser system failure.
A nozzle-riser system failure could also generate the coincident failure of a jet pump diffuser body; however, the converse is not true.
3.6.F/4.6.F     Recirculation Pum   0 eratio Steady-state   operation without forced recirculation     will not be permitted for more than 12 hours.     And the start of   a recirculation   pump from the natural circulation condition will not be permitted unless the temperature difference between the loop to be started and the core coolant temperature is less than 75 F. This reduces the positive reactivity insertion to an acceptably low value.
The lack of any substantial stress in the jet pump diffuser body makes failure impossible without an initial nozzle-riser system failure.
3.6.F/4.6.F Recirculation Pum 0 eratio Steady-state operation without forced recirculation will not be permitted for more than 12 hours.
And the start of a recirculation pump from the natural circulation condition will not be permitted unless the temperature difference between the loop to be started and the core coolant temperature is less than 75 F.
This reduces the positive reactivity insertion to an acceptably low value.
Requiring the discharge valve of the lower speed loop to remain closed until the speed of the faster pump is below 50 percent of its rated speed provides assurance when going from one-to-two pump operation that excessive vibration of the jet pump risers will not occur.
Requiring the discharge valve of the lower speed loop to remain closed until the speed of the faster pump is below 50 percent of its rated speed provides assurance when going from one-to-two pump operation that excessive vibration of the jet pump risers will not occur.
3.6.G/4.6.G     Structural Inte   it The requirements   for the reactor coolant systems inservice inspection program have been identified by evaluating the need for a sampling examination of areas of high stress and highest probability of failure in the system and the need to meet as closely as possible the requirements of Section XI, of the ASME Boiler and Pressure Vessel Code.
3.6.G/4.6.G Structural Inte it The requirements for the reactor coolant systems inservice inspection program have been identified by evaluating the need for a sampling examination of areas of high stress and highest probability of failure in the system and the need to meet as closely as possible the requirements of Section XI, of the ASME Boiler and Pressure Vessel Code.
The program   reflects the built-in limitations of     access to the reactor coolant systems.
The program reflects the built-in limitations of access to the reactor coolant systems.
It is   intended that the required examinations and inspection be completed during each 10-year interval. The periodic examinations are to be done during refueling outages or other extended plant shutdown periods.
It is intended that the required examinations and inspection be completed during each 10-year interval.
BFN                                     3.6/4.6-32 Unit  3
The periodic examinations are to be done during refueling outages or other extended plant shutdown periods.
BFN Unit 3 3.6/4.6-32


3.6/4.6 BASES 3.6.G/4.6.G (Cont'd)
3.6/4.6 BASES 3.6.G/4.6.G (Cont'd)
Only proven nondestructive   testing techniques   will be used.
Only proven nondestructive testing techniques will be used.
More frequent inspections shall be performed on certain circumferential pipe welds as   listed in Section 4.6.G.4 to provide additional protection against pipe whip. These welds were selected in respect to their distance from hangers or supports wherein a failure of the weld would permit the unsupported segments of pipe to strike the drywell wall or nearby auxiliary systems or control systems. Selection was based on judgment from actual plant observation of hanger and support locations and review of drawings.
More frequent inspections shall be performed on certain circumferential pipe welds as listed in Section 4.6.G.4 to provide additional protection against pipe whip.
Inspection of all these welds during each 10-year inspection interval will result in three additional examinations above the requirements of Section XI of ASME Code.
These welds were selected in respect to their distance from hangers or supports wherein a failure of the weld would permit the unsupported segments of pipe to strike the drywell wall or nearby auxiliary systems or control systems.
References
Selection was based on judgment from actual plant observation of hanger and support locations and review of drawings.
: l. Inservice Inspection   and Testing (BFNP FSAR Subsection 4.12)
Inspection of all these welds during each 10-year inspection interval will result in three additional examinations above the requirements of Section XI of ASME Code.
: 2. Inservice Inspection of Nuclear Reactor Coolant Systems, Section XI,     ASME Boiler and Pressure Vessel Code
References l.
: 3. ASME Boiler and Pressure Vessel Code, Section   III (1968 Edition)
Inservice Inspection and Testing (BFNP FSAR Subsection 4.12) 2.
: 4. American Society for Nondestructive Testing   No. SNT-TC-1A (1968   Edition)
Inservice Inspection of Nuclear Reactor Coolant Systems, Section XI, ASME Boiler and Pressure Vessel Code 3.
BFN                                     3.6/4.6-33 Unit  3
ASME Boiler and Pressure Vessel Code, Section III (1968 Edition) 4.
American Society for Nondestructive Testing No. SNT-TC-1A (1968 Edition)
BFN Unit 3 3.6/4.6-33


Enclosure   2 REASON FOR CHANGES, DESCRIPTION, AND JUSTIFICATION BROWNS FERRY NUCLEAR PLANT (BFN)
Enclosure 2
EASO   FOR CHA GES BFN units 1, 2, and 3 technical specifications (TSs) are being changed to:
REASON FOR CHANGES, DESCRIPTION, AND JUSTIFICATION BROWNS FERRY NUCLEAR PLANT (BFN)
(1) revise Table 3.2.B and Limiting Conditions for Operation (LCO) 3.5.B.ll, 3.5.E.1, 3.5.F.1, 3.5.G.l, and 3.6.D.l and the bases section for 3.6.D/4.6.D to clarify equipment operability requirements when the reactor is in the cold shutdown condition, (2) revise the maximum operating power level allowed with an inoperable Recirculation Pump Trip (RPT) system(s) from 85 percent to 30 percent power, and (3) correct two typographical errors in Table 3.2.B.
EASO FOR CHA GES BFN units 1, 2, and 3 technical specifications (TSs) are being changed to:
ESCRIPTIO   A D JUSTIFIC TIO     OR   E   0 OS     C   G S
(1) revise Table 3.2.B and Limiting Conditions for Operation (LCO) 3.5.B.ll, 3.5.E.1, 3.5.F.1, 3.5.G.l, and 3.6.D.l and the bases section for 3.6.D/4.6.D to clarify equipment operability requirements when the reactor is in the cold shutdown condition, (2) revise the maximum operating power level allowed with an inoperable Recirculation Pump Trip (RPT) system(s) from 85 percent to 30 percent
: 1. Clarify equipment   operability requirements for       when the reactor is in the cold shutdown condition.
: power, and (3) correct two typographical errors in Table 3.2.B.
: a. Add the   following note to Table 3.2.B for the drywell high pressure instruments (PIS-64-57 A-D, PIS-64-58 A-D, and PIS-64-58 E-H for unit 2 and PS-64-57 A-D, PS-64-58 A-D, and PS-64-58 E-H           for units   1 and 3):
ESCRIPTIO A
            "18. Not required to be   OPERABLE   in the COLD SHUTDOWN CONDITION."
D JUSTIFIC TIO OR E
: b. Existing LCO 3.5.B.ll reads in part: "When there is irradiated fuel in the reactor and the reactor vessel pressure is greater than atmospheric, Revised   LCO 3.5.B.ll would read   in part: "When there is irradiated fuel in the reactor     and the reactor is not in the COLD SHUTDOWN CONDITION,
0 OS C
: c. Existing   LCO 3.5.E.1   reads in part: "The HPCI system shall         be OPERABLE . . . except     as specified in Specification 3.5.E.2 Revised   LCO 3.5.E.1 would read     in part: "The HPCI system     shall   be OPERABLE . . . except     in the COLD SHUTDOWN CONDITION or as       specified in 3.5.E.2
G S
: d. Existing   LCO 3.5.F.l reads in part:     "The RCICS shall be OPERABLE except as specified     in 3.5.F.2.
1.
Revised LCO 3.5.F.1 would read in part: "The RCICS shall be operable except in the COLD SHUTDOWN CONDITION or as specified in 3.5.F.2
Clarify equipment operability requirements for when the reactor is in the cold shutdown condition.
: e. Existing LCO 3.5.G.1 reads in part: "Four of the six valves of the Automatic Depressurization System shall be OPERABLE . . . except as specified in 3.5.G.2 and 3.5.G.3 below."
a.
Revised LCO 3.5.G.1 would read in part: "Four           of the six valves of the Automatic Depressurization System shall be         OPERABLE .  .  . except in the   COLD SHUTDOWN CONDITION   or as specified in 3.5.G.2 or 3.5.G.3 below."
Add the following note to Table 3.2.B for the drywell high pressure instruments (PIS-64-57 A-D, PIS-64-58 A-D, and PIS-64-58 E-H for unit 2 and PS-64-57 A-D, PS-64-58 A-D, and PS-64-58 E-H for units 1 and 3):
"18.
Not required to be OPERABLE in the COLD SHUTDOWN CONDITION."
b.
Existing LCO 3.5.B.ll reads in part:
"When there is irradiated fuel in the reactor and the reactor vessel pressure is greater than atmospheric, Revised LCO 3.5.B.ll would read in part:
"When there is irradiated fuel in the reactor and the reactor is not in the COLD SHUTDOWN CONDITION, c.
Existing LCO 3.5.E.1 reads in part:
"The HPCI system shall be OPERABLE
. except as specified in Specification 3.5.E.2 Revised LCO 3.5.E.1 would read in part:
"The HPCI system shall be OPERABLE
. except in the COLD SHUTDOWN CONDITION or as specified in 3.5.E.2 d.
Existing LCO 3.5.F.l reads in part:
"The RCICS shall be OPERABLE except as specified in 3.5.F.2.
Revised LCO 3.5.F.1 would read in part:
"The RCICS shall be operable except in the COLD SHUTDOWN CONDITION or as specified in 3.5.F.2 e.
Existing LCO 3.5.G.1 reads in part:
"Four of the six valves of the Automatic Depressurization System shall be OPERABLE
. except as specified in 3.5.G.2 and 3.5.G.3 below."
Revised LCO 3.5.G.1 would read in part:
"Four of the six valves of the Automatic Depressurization System shall be OPERABLE
. except in the COLD SHUTDOWN CONDITION or as specified in 3.5.G.2 or 3.5.G.3 below."


Page 2   of 5
Page 2 of 5 f.
: f. Existing     LCO 3.6.D.1 reads in part:   "When more than one   relief valves are known to be     failed Revised LCO   3.6.D.l would read in part:   "When more   than one relief valve is known to be failed
Existing LCO 3.6.D.1 reads in part:
: g. Add the following to LCO 3.6.D.1:     "The relief valves are   not required to be OPERABLE in the COLD SHUTDOWN CONDITION."
"When more than one relief valves are known to be failed Revised LCO 3.6.D.l would read in part:
: h. Add the following paragraph to the bases section for 3.6.D/4.6.D:
"When more than one relief valve is known to be failed g.
    "The relief valves are not required to be OPERABLE in the COLD SHUTDOWN CONDITION. Overpressure protection is provided during hydrostatic tests by two of the relief valves whose relief setting has been established in conformance with ASME section XI code requirements. The capacity of one relief valve exceeds the charging capacity of the pressurization source used during hydrostatic testing. Two relief valves are used to provide redundancy."
Add the following to LCO 3.6.D.1:
Table 3.2.B currently requires the drywell high pressure instruments to be operable whenever any core and containment cooling system in TS section 3.5 is operable. Change (a) above would allow these instruments to be inoperable when the plant is in the cold shutdown condition. Initiation of these instruments, along with the low reactor pressure instruments, indicates a breach of the nuclear system process barrier within the drywell (steam leak). With the reactor in the cold shutdown condition (reactor coolant temperature g 212'F and reactor in shutdown or refuel mode), there is no need to detect steam leaks so     it is acceptable for the drywell high pressure instruments to be inoperable.
"The relief valves are not required to be OPERABLE in the COLD SHUTDOWN CONDITION."
Additionally,   TVA is required by TS section 4.7.A to conduct an Integrated Leakrate Test (ILRT) at certain frequencies. The ILRT demonstrates the ability of the primary containment to contain the gases released from the primary system during a postulated worst case accident with leakage rates low enough to ensure exposure rates less than the 10 CFR 100 limits. The test is performed with,the plant in the cold shutdown condition by pressurizing the primary containment (drywell and torus) to design bases accident pressure (49.6 psig) and monitoring pressure and temperature for a prescribed period of time. From this data, the leakage can be calculated.
h.
The high drywell pressure instruments listed above have a trip level setting of   between 1 and 2.5 psig. Inhibiting these pressure instruments during the ILRT is required to prevent unnecessary Emergency Core Cooling System (ECCS)   initiations.
Add the following paragraph to the bases section for 3.6.D/4.6.D:
The Residual Heat Removal (RHR) and Core Spray (CS) systems are required to be operable during the test in accordance with LCOs 3.5.A and 3.5.B because reactor pressure is greater than atmospheric.       The reactor low water level instruments (LS-3-58 A-D) are operable and initiate the RHR         or CS systems on a low-low reactor water level     if necessary. This ensures that RHR and CS could provide sufficient makeup capacity       if required to protect the fuel.
"The relief valves are not required to be OPERABLE in the COLD SHUTDOWN CONDITION.
Overpressure protection is provided during hydrostatic tests by two of the relief valves whose relief setting has been established in conformance with ASME section XI code requirements.
The capacity of one relief valve exceeds the charging capacity of the pressurization source used during hydrostatic testing.
Two relief valves are used to provide redundancy."
Table 3.2.B currently requires the drywell high pressure instruments to be operable whenever any core and containment cooling system in TS section 3.5 is operable.
Change (a) above would allow these instruments to be inoperable when the plant is in the cold shutdown condition.
Initiation of these instruments, along with the low reactor pressure instruments, indicates a breach of the nuclear system process barrier within the drywell (steam leak).
With the reactor in the cold shutdown condition (reactor coolant temperature g 212'F and reactor in shutdown or refuel mode),
there is no need to detect steam leaks so it is acceptable for the drywell high pressure instruments to be inoperable.
Additionally, TVA is required by TS section 4.7.A to conduct an Integrated Leakrate Test (ILRT) at certain frequencies.
The ILRT demonstrates the ability of the primary containment to contain the gases released from the primary system during a postulated worst case accident with leakage rates low enough to ensure exposure rates less than the 10 CFR 100 limits.
The test is performed with,the plant in the cold shutdown condition by pressurizing the primary containment (drywell and torus) to design bases accident pressure (49.6 psig) and monitoring pressure and temperature for a prescribed period of time.
From this data, the leakage can be calculated.
The high drywell pressure instruments listed above have a trip level setting of between 1 and 2.5 psig.
Inhibiting these pressure instruments during the ILRT is required to prevent unnecessary Emergency Core Cooling System (ECCS) initiations.
The Residual Heat Removal (RHR) and Core Spray (CS) systems are required to be operable during the test in accordance with LCOs 3.5.A and 3.5.B because reactor pressure is greater than atmospheric.
The reactor low water level instruments (LS-3-58 A-D) are operable and initiate the RHR or CS systems on a low-low reactor water level if necessary.
This ensures that RHR and CS could provide sufficient makeup capacity if required to protect the fuel.


Page 3 of 5 The standby gas treatment and secondary containment systems     are also operable during the test and available to contain and     filter any radioactive material were   it to be released.
Page 3 of 5 The standby gas treatment and secondary containment systems are also operable during the test and available to contain and filter any radioactive material were it to be released.
TSs currently require that the RHR crosstie be operable with reactor pressure greater than atmospheric (LCO 3.5.B.ll), the High Pressure Coolant Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) systems must be operable with reactor pressure greater the 150 psig, and the Automatic Depressurization System (ADS) and relief valves must be operable with reactor pressure greater than 105 psig.
TSs currently require that the RHR crosstie be operable with reactor pressure greater than atmospheric (LCO 3.5.B.ll), the High Pressure Coolant Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) systems must be operable with reactor pressure greater the 150 psig, and the Automatic Depressurization System (ADS) and relief valves must be operable with reactor pressure greater than 105 psig.
With the reactor in the cold shutdown condition, vessel pressure is atmospheric except during performance of inservice hydrostatic pressure tests, inservice leakage tests   and ILRT.
With the reactor in the cold shutdown condition, vessel pressure is atmospheric except during performance of inservice hydrostatic pressure
Once per inservice inspection interval, the plant is required to perform an inservice hydrostatic pressure test on the reactor vessel and attached piping out to and including the first isolation valve to ensure the retaining capability of the reactor coolant pressure boundary. The test is performed at pressures (1096 to 1150 psia in the dome) in excess of normal operating pressure (approximately 1020 psia).
: tests, inservice leakage tests and ILRT.
An inservice leakage test is requi.red whenever the reactor coolant pressure retaining boundary is breached. This test is similar to the hydrostatic test, but   it is performed at normal operating pressure (approximately 1020 psia). As described previously, the ILRT is performed by pressurizing the primary containment to 49.6 psig.
Once per inservice inspection interval, the plant is required to perform an inservice hydrostatic pressure test on the reactor vessel and attached piping out to and including the first isolation valve to ensure the retaining capability of the reactor coolant pressure boundary.
As currently written, TSs require the RHR crosstie to be operable for each of these tests and HPCI, RCIC, ADS, and relief valves to be operable for the inservice hydrostatic and leakage tests. In reality, the ADS is disabled and HPCI and RCIC, both steam turbine driven systems, have no steam supply available during these tests.
The test is performed at pressures (1096 to 1150 psia in the dome) in excess of normal operating pressure (approximately 1020 psia).
These tests are performed in the cold shutdown condition at the end of the refueling outage with fuel loaded and the reactor pressure vessel head installed. These tests occur when primary system energy is minimal with all control rods inserted. Because the reactor vessel pressure is greater than atmospheric the   RHR and CS systems are required to be operable.
An inservice leakage test is requi.red whenever the reactor coolant pressure retaining boundary is breached.
The inservice hydrostatic and leakage tests are performed at or above a minimum temperature as specified by TS figure 3.6-1. With the system temperature (approximately 207'F) below the atmospheric pressure boiling point, enthalpy of the bulk fluid is low. If a leak greater than the makeup capacity of the Control Rod Drive (CRD) pump should occur during the test, the system would be depressurized well below the maximum pressure at which the RHR and CS systems could inject to the vessel before water level dropped to an unsafe level. The available RHR and CS systems are sufficient to preclude fuel uncovering in the event of a leak.
This test is similar to the hydrostatic test, but it is performed at normal operating pressure (approximately 1020 psia).
As described previously, the ILRT is performed by pressurizing the primary containment to 49.6 psig.
As currently written, TSs require the RHR crosstie to be operable for each of these tests and HPCI, RCIC, ADS, and relief valves to be operable for the inservice hydrostatic and leakage tests.
In reality, the ADS is disabled and HPCI and RCIC, both steam turbine driven systems, have no steam supply available during these tests.
These tests are performed in the cold shutdown condition at the end of the refueling outage with fuel loaded and the reactor pressure vessel head installed.
These tests occur when primary system energy is minimal with all control rods inserted.
Because the reactor vessel pressure is greater than atmospheric the RHR and CS systems are required to be operable.
The inservice hydrostatic and leakage tests are performed at or above a
minimum temperature as specified by TS figure 3.6-1.
With the system temperature (approximately 207'F) below the atmospheric pressure boiling point, enthalpy of the bulk fluid is low. If a leak greater than the makeup capacity of the Control Rod Drive (CRD) pump should occur during the test, the system would be depressurized well below the maximum pressure at which the RHR and CS systems could inject to the vessel before water level dropped to an unsafe level.
The available RHR and CS systems are sufficient to preclude fuel uncovering in the event of a leak.


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Page 4 of 5 The RHR crosstie is provided to maintain a long term reactor core and primary containment cooling capability irrespective of primary containment integrity or operability of the RHR system on a given unit. This is provided in case the torus is breached, flooding the RHR pumps on the affected unit. However, with the reactor in the cold shutdown condition, there is no high energy potential to breach the torus so the RHR crosstie is not needed.
Page 4 of 5 The RHR crosstie is provided to maintain a long term reactor core and primary containment cooling capability irrespective of primary containment integrity or operability of the RHR system on a given unit.
During inservice hydrostatic testing,     ll of the 13 relief valves are disabled by removing the pilot cartridges and blanking the pilot ports.
This is provided in case the torus is breached, flooding the RHR pumps on the affected unit.
However, with the reactor in the cold shutdown condition, there is no high energy potential to breach the torus so the RHR crosstie is not needed.
During inservice hydrostatic testing, ll of the 13 relief valves are disabled by removing the pilot cartridges and blanking the pilot ports.
Overpressure protection is provided by the two remaining relief valves which have their setpoint established in accordance with ASME Section XI.
Overpressure protection is provided by the two remaining relief valves which have their setpoint established in accordance with ASME Section XI.
The relief capacity of one relief valve exceeds the flow capacity of the hydrostatic pressure source. Two valves are used for redundancy.
The relief capacity of one relief valve exceeds the flow capacity of the hydrostatic pressure source.
These changes   are consistent with the General Electric Boiling Water Reactor Standard TSs (NUREG 0123) which requires HPCI (section 3.5.1.c),
Two valves are used for redundancy.
RCIC (section 3.7.4), ADS (section 3.5.1.d), and relief valves (section 3.4.2.1) to be operable only in the power operation, startup, and hot shutdown conditions.
These changes are consistent with the General Electric Boiling Water Reactor Standard TSs (NUREG 0123) which requires HPCI (section 3.5.1.c),
: 2. Correct the   maximum power level allowed with an inoperable RPT system(s).
RCIC (section 3.7.4),
Existing Table 3.2.B, Note   17 reads:
ADS (section 3.5.1.d),
  "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 1 RPT system is inoperable for more than 72 hours, an orderly power reduction shall be initiated and reactor power shall be less than 85 percent within four hours."
and relief valves (section 3.4.2.1) to be operable only in the power operation,
: startup, and hot shutdown conditions.
2.
Correct the maximum power level allowed with an inoperable RPT system(s).
Existing Table 3.2.B, Note 17 reads:
"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 1 RPT system is inoperable for more than 72 hours, an orderly power reduction shall be initiated and reactor power shall be less than 85 percent within four hours."
Proposed change to Table 3.2.B, Note 17 would read:
Proposed change to Table 3.2.B, Note 17 would read:
  "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 hours, an orderly power reduction shall be initiated and reactor power shall be less than 30 percent within four hours."
"17.
This change corrects the maximum operating power level allowed with an inoperable RPT system(s) from 85 percent to 30 percent Core Thermal Power (CTP). Thirty percent CTP is used in the RPT analysis (NED0-24119, "Basis for Installation of Recirculation Pump Trip System for Browns Ferry,"
Two RPT systems exist, either of which will trip both recirculation pumps.
April 1978, BFN Updated Final Safety Analysis Report (UFSAR) Section 7.9.4.5), and is conservatively determined to be the maximum power level at which fuel cladding integrity can be assumed during an end of cycle limiting overpressurization event without RPT protection.
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 hours, an orderly power reduction shall be initiated and reactor power shall be less than 30 percent within four hours."
This change corrects the maximum operating power level allowed with an inoperable RPT system(s) from 85 percent to 30 percent Core Thermal Power (CTP).
Thirty percent CTP is used in the RPT analysis (NED0-24119, "Basis for Installation of Recirculation Pump Trip System for Browns Ferry,"
April 1978, BFN Updated Final Safety Analysis Report (UFSAR) Section 7.9.4.5),
and is conservatively determined to be the maximum power level at which fuel cladding integrity can be assumed during an end of cycle limiting overpressurization event without RPT protection.


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Page 5 of 5 The RPT   provides automatic trip of both recirculation pumps after a turbine trip or generator load rejection     if reactor power is above approximately   30 percent of rated full load. The purpose of this trip is to reduce the peak reactor pressure and peak heat flux resulting from transients in which   it is postulated that there is a coincident failure of the turbine bypass system. The recirculation pump trip signal results from either turbine control valve fast closure or turbine stop valve closure. Reactor scram is also initiated by these signals. The very rapid reduction in core flow following a recirculation pump trip early in
Page 5 of 5 The RPT provides automatic trip of both recirculation pumps after a turbine trip or generator load rejection if reactor power is above approximately 30 percent of rated full load.
  ,these transients reduces the severity of these events because the immediate resultant increase in core voids provides negative reactivity which supplements the negative reactivity from control rod scram.
The purpose of this trip is to reduce the peak reactor pressure and peak heat flux resulting from transients in which it is postulated that there is a coincident failure of the turbine bypass system.
The proposed   change reduces the trip set point from 85 percent to 30 percent and   is therefore   more conservative than the current operational requirements.
The recirculation pump trip signal results from either turbine control valve fast closure or turbine stop valve closure.
Additionally, the   number "1" in Note 17 is being revised to the alphabetic "one" to be consistent with the rest of the note.
Reactor scram is also initiated by these signals.
: 3. Correct two typographical errors in Table 3.2.B.
The very rapid reduction in core flow following a recirculation pump trip early in
: a. Correct typographical error in the     first entry under "Remarks" in Table 3.2.B (Page 3.2/4.2-14).
,these transients reduces the severity of these events because the immediate resultant increase in core voids provides negative reactivity which supplements the negative reactivity from control rod scram.
The proposed change reduces the trip set point from 85 percent to 30 percent and is therefore more conservative than the current operational requirements.
Additionally, the number "1" in Note 17 is being revised to the alphabetic "one" to be consistent with the rest of the note.
3.
Correct two typographical errors in Table 3.2.B.
a.
Correct typographical error in the first entry under "Remarks" in Table 3.2.B (Page 3.2/4.2-14).
Existing entry reads:
Existing entry reads:
        "1. Below trip setting initiated   HPCI."
"1.
Below trip setting initiated HPCI."
Proposed change to Table 3.2.B would read:
Proposed change to Table 3.2.B would read:
        "1. Below trip setting initiates   HPCI."
"1.
This change revises the word     "initiated" to "initiates" so that this entry will be in the present tense     like the other remarks in Table 3.2.B.
Below trip setting initiates HPCI."
: b. Correct typographical error under "Minimum No. Operable Per Trip Sys" column on Table 3.2.B for "RHR (LPCI) Trip System bus power monitor" (page 3.2/4.2-17) Unit 2 only.
This change revises the word "initiated" to "initiates" so that this entry will be in the present tense like the other remarks in Table 3.2.B.
The entry in this column should be   "l."
b.
This is an omission in the Unit 2 TSs. The original BFN Unit 2       TSs indicate a "1" in this column as do the current BFN Units 1 and       3 TSs.
Correct typographical error under "Minimum No. Operable Per Trip Sys" column on Table 3.2.B for "RHR (LPCI) Trip System bus power monitor" (page 3.2/4.2-17) Unit 2 only.
The entry in this column should be "l."
This is an omission in the Unit 2 TSs.
The original BFN Unit 2 TSs indicate a "1" in this column as do the current BFN Units 1 and 3 TSs.


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PROPOSED   DETERINATION OF NO SIGNIFICANT HAZARDS BROWNS FERRY NUCLEAR PLANT (BFN)
PROPOSED DETERINATION OF NO SIGNIFICANT HAZARDS BROWNS FERRY NUCLEAR PLANT (BFN)
DESCRIPTIO   OF PRO OSE   TECH ICAL SPECIFICATION   S A   D   T BFN units 1, 2, and 3 technical specifications (TSs) are being changed to:
DESCRIPTIO OF PRO OSE TECH ICAL SPECIFICATION S
(1) revise Table 3.2.B and Limiting Conditions for Operation (LCO) 3.5.B.11, 3.5.E.1, 3.5.F.l, 3.5.G.l, and 3.6.D.1 and the bases section for 3.6.D/4.6.D to clarify equipment operability requirements when the reactor is in the cold shutdown condition, (2) revise the maximum operating power level allowed with an inoperable Recirculation Pump Trip (RPT) system(s) from 85 percent to 30 percent power, and (3) correct two typographical errors in Table 3.2.B.
A D
B SIS FOR PROPOSED   0 SIG I ICA   H   RDS CONSIDERATIO DE ERM   TIO NRC has provided standards for determining whether a significant hazards consideration exists as stated in 10 CFR 50.92(c). A proposed amendment, to, an operating license involves no significant hazards considerations     if operation of the facility in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated, or (2) create the possibility of a new or different kind of accident from an accident previously evaluated, or (3) involve a significant reduction in a margin of safety.
T BFN units 1, 2, and 3 technical specifications (TSs) are being changed to:
: 1. The proposed changes do not   involve a significant increase in the probability or consequences   of an accident previously evaluated. Change 1 clarifies equipment operability requirements with the reactor in the cold shutdown condition. With the reactor in the cold shutdown condition, primary system energy is minimal and the control rods are inserted.
(1) revise Table 3.2.B and Limiting Conditions for Operation (LCO) 3.5.B.11, 3.5.E.1, 3.5.F.l, 3.5.G.l, and 3.6.D.1 and the bases section for 3.6.D/4.6.D to clarify equipment operability requirements when the reactor is in the cold shutdown condition, (2) revise the maximum operating power level allowed with an inoperable Recirculation Pump Trip (RPT) system(s) from 85 percent to 30 percent
Reactor pressure is normally atmospheric except during performance of inservice hydrostatic tests, inservice leakage tests, and Integrated Leak Rate Tests (ILRT). This change would inhibit the drywell high pressure instruments which function to detect primary system leaks. With minimal system energy and no steam generation, this function is not required. The High Pressure   Coolant Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) systems are not required because there is no steam supply to operate them and Residual Heat Removal (RHR) and Core Spray (CS) are operable and capable of providing makeup in case of leaks to protect the fuel from being uncovered. The Automatic Depressurization System (ADS) is not required for leaks considered possible during the inservice hydrostatic test. Reactor pressure would decrease fast enough to allow residual heat removal and core spray injection in time to preclude water level decreasing to an unsafe level. The relief valves are not required to be operable because alternate means of overpressurization protection are provided in the tests. During inservice hydrostatic testing,       ll of the 13 relief valves are disabled by removing the pilot cartridges and blanking the pilot ports. Overpressure protection is provided by the two remaining relief valves which have their setpoint established in accordance with ASME Section XI. The RHR crosstie is not required because there is no high energy potential to breach the torus in the cold shutdown condition. The change is consistent with industry practice and the GE BWR Standard TSs   (NUREG 0123).
: power, and (3) correct two typographical errors in Table 3.2.B.
B SIS FOR PROPOSED 0 SIG I ICA H
RDS CONSIDERATIO DE ERM TIO NRC has provided standards for determining whether a significant hazards consideration exists as stated in 10 CFR 50.92(c).
A proposed amendment, to, an operating license involves no significant hazards considerations if operation of the facility in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated, or (2) create the possibility of a new or different kind of accident from an accident previously evaluated, or (3) involve a significant reduction in a margin of safety.
1.
The proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.
Change 1
clarifies equipment operability requirements with the reactor in the cold shutdown condition.
With the reactor in the cold shutdown condition, primary system energy is minimal and the control rods are inserted.
Reactor pressure is normally atmospheric except during performance of inservice hydrostatic tests, inservice leakage tests, and Integrated Leak Rate Tests (ILRT).
This change would inhibit the drywell high pressure instruments which function to detect primary system leaks.
With minimal system energy and no steam generation, this function is not required.
The High Pressure Coolant Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) systems are not required because there is no steam supply to operate them and Residual Heat Removal (RHR) and Core Spray (CS) are operable and capable of providing makeup in case of leaks to protect the fuel from being uncovered.
The Automatic Depressurization System (ADS) is not required for leaks considered possible during the inservice hydrostatic test.
Reactor pressure would decrease fast enough to allow residual heat removal and core spray injection in time to preclude water level decreasing to an unsafe level.
The relief valves are not required to be operable because alternate means of overpressurization protection are provided in the tests.
During inservice hydrostatic testing, ll of the 13 relief valves are disabled by removing the pilot cartridges and blanking the pilot ports.
Overpressure protection is provided by the two remaining relief valves which have their setpoint established in accordance with ASME Section XI.
The RHR crosstie is not required because there is no high energy potential to breach the torus in the cold shutdown condition.
The change is consistent with industry practice and the GE BWR Standard TSs (NUREG 0123).


Page 2   of 2 Change 2   is a more conservative requirement. The RPT system provides an automatic trip of both recirculation pumps after a turbine trip or a generator load reject. This reduction in flow increases the core voids and provides immediate negative reactivity to reduce the severity of the transient. There are two RPT systems.         If both RPT systems are inoperable or if one RPT system is inoperable for more than 72 hours, reactor power shall be less than 30 percent within four hours (vs. the current 85 percent). The proposed value of 30 percent power is consistent with the BFN RPT analysis and the BFN Updated Final Safety Analysis Report.
Page 2 of 2 Change 2 is a more conservative requirement.
The RPT system provides an automatic trip of both recirculation pumps after a turbine trip or a generator load reject.
This reduction in flow increases the core voids and provides immediate negative reactivity to reduce the severity of the transient.
There are two RPT systems.
If both RPT systems are inoperable or if one RPT system is inoperable for more than 72 hours, reactor power shall be less than 30 percent within four hours (vs. the current 85 percent).
The proposed value of 30 percent power is consistent with the BFN RPT analysis and the BFN Updated Final Safety Analysis Report.
Therefore, this change involves no significant increase in the probability or consequences of an accident previously analyzed.
Therefore, this change involves no significant increase in the probability or consequences of an accident previously analyzed.
Change 3   is an administrative   change that corrects typographical errors.
Change 3 is an administrative change that corrects typographical errors.
: 2. The proposed   change does not create the possibility of a new or different kind of accident from an accident previously evaluated. Change             l does not involve changes in plant hardware or method of operation from that currently practiced. The changes are clarifications to TSs to facilitate performance of required TS testing with the reactor in the cold shutdown condition. The methods of performance are consistent with industry practice.
2.
Change 2   will ensure   that when both RPT systems are inoperable or when one RPT system   is inoperable   more than 72 hours, reactor power is dropped to a level consistent with the analysis performed for the         RPT installation.
The proposed change does not create the possibility of a new or different kind of accident from an accident previously evaluated.
Change 3   corrects   two typographical errors so the   TSs will be   more consistent.
Change l does not involve changes in plant hardware or method of operation from that currently practiced.
: 3. The proposed   changes do not involve a significant reduction in the margin of safety. Change l clarifies equipment operability requirements with the reactor in the cold shutdown condition. Sufficient safety equipment is still available to ensure the fuel remains covered, even in the event of leaks. It does not reduce the equipment available to mitigate an accident and as such does not reduce the margin of safety.
The changes are clarifications to TSs to facilitate performance of required TS testing with the reactor in the cold shutdown condition.
Change 2   is more conservative than the     current TS. When the   RPT system   is inoperable the maximum allowed reactor       power will be   reduced. This is consistent with the analysis performed       for the RPT installation and the FSAR and does not reduce the margin of       safety.
The methods of performance are consistent with industry practice.
Change 3   is an administrative   change which does not reduce the margin       of safety.
Change 2 will ensure that when both RPT systems are inoperable or when one RPT system is inoperable more than 72 hours, reactor power is dropped to a level consistent with the analysis performed for the RPT installation.
These changes have been reviewed by TVA. Based on this review, TVA does not believe the changes present a Significant Hazards Consideration.
Change 3 corrects two typographical errors so the TSs will be more consistent.
3.
The proposed changes do not involve a significant reduction in the margin of safety.
Change l clarifies equipment operability requirements with the reactor in the cold shutdown condition.
Sufficient safety equipment is still available to ensure the fuel remains
: covered, even in the event of leaks.
It does not reduce the equipment available to mitigate an accident and as such does not reduce the margin of safety.
Change 2 is more conservative than the current TS.
When the RPT system is inoperable the maximum allowed reactor power will be reduced.
This is consistent with the analysis performed for the RPT installation and the FSAR and does not reduce the margin of safety.
Change 3 is an administrative change which does not reduce the margin of safety.
These changes have been reviewed by TVA.
Based on this review, TVA does not believe the changes present a Significant Hazards Consideration.


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Latest revision as of 01:51, 7 January 2025

Proposed Tech Specs Changes to Table 3.2.B & LCOs 3.5.B.11 & 3.5.E.1
ML18033B541
Person / Time
Site: Browns Ferry  Tennessee Valley Authority icon.png
Issue date: 10/30/1990
From:
TENNESSEE VALLEY AUTHORITY
To:
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ML18033B540 List:
References
NUDOCS 9011090356
Download: ML18033B541 (80)
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Text

UNIT 1 EFFECTIVE PAGE LIST REMOVE INSERT 3.2/4.2-14 3.2/4.2-15 3.2/4.2-23 3.2/4.2-24 3.5/4.5-7 3.5/4.5-8 3.5/4.5-12 3.5/4.5-13 3.5/4.5-14 3.5/4.5-15 3.5/4.5-16 3.5/4.5-17 3.6/4.6-9 3.6/4.6-10 3.6/4.6-30 3.6/4.6-31 3.6/4.6-32 3.6/4.6-33 3.2/4.2-14 3.2/4.2-15 3.2/4.2-23*

3.2/4.2-24 3.5/4.5-7 3.5/4.5-8*

3.5/4.5-12*

3.5/4.5-13 3.5/4.5-14 3.5/4.5-15*

3.5/4.5-16 3.5/4.5-17*

3.6/4.6-9*

3.6/4.6-10 3.6/4.6-30*

3.6/4.6-31 3.6/4.6-32*

3.6/4.6-33*

  • Denotes overleaf or spillover page.

09035b 901030 PDR ADOCK 05000>59 PNU

TABLE 3.2.8 INSTRUHENTATION THAT INITATES OR CONTROLS THE CORE AND CONTAINHENT COOLING SYSTEHS Hinimum No.

Operable Per

~Tri )~s~l 2(16) 1(16)

Functi n

Instrument Channel Reactor Low Water Level Instrument Channel Reactor Low Water Level Instrument Channel Reactor Low Water Level (LIS-3-58A-D, SW ¹1)

Instrument Channel-Reactor Low Mater Level (LIS-3-58A-D, SW ¹2)

Instrument Channel Reactor Low Water Level Permissive (LIS-3-184 8

185, SW ¹1)

Instrument Channel Reactor Low Water Level (LITS-3-52 and 62, SW ¹1)

Tri level S t in

~Ac icn

> 470" above vessel zero

> 470" above vessel zero.

> 378" above vessel zero.

> 378" above vessel zero.

> 544" above vessel zero.

A

> 312 5/16" above vessel zero.

A (2/3 core height)

R mark 1.

Below trip setting initiates HPCI.

1.

Hultiplier relays initiate RCIC.

1.

Below trip setting initiates CSS.

Hultiplier relays initiate LPCI.

2.

Hultiplier relay from CSS initiates accident signal (15).

1.

Below trip settings, in conjunction with drywell high pressure, low water level permissive, 120 sec.

delay timer and CSS or RHR pump running, initiates ADS.

l.

Below trip setting permissive for initiating si gnal s on ADS.

l.

Below trip setting prevents inadvertent operation of containment spray during accident condition.

4 4

Sr,

>~

4 Hinimum No.

Operable Per

~Tri S

i 2(18) 2(18) 2(18) 2(16)(18)

Function Instrument Channel-Drywell High Pressure (PS-64-58 E-H)

Instrument Channel-Drywell High Pressure (PS-64-58 A-D, SW ¹2)

Instrument Channel-Drywell High Pressure (PS-64-58A-D, SW ¹1)-

Instrument Channel-Drywell High Pressure (PS-64-57A-0)

TABLE 3.2.8 (Continued)

Tri L v 1

in 1< p<2.5 psig

< 2.5 psig

< 2.5 psig

< 2.5 psig Action Remark l.

Below trip setting prevents inadvertent operation of containment spray during accident conditions.

l.

Above trip setting in con-junction with low reactor pressure initiates CSS.

Hultiplier relays initiate HPCI.

2.

Hultiplier relay from CSS initiates ace i dent s ignal. (15) l.

Above trip setting in conjunction with low reactor pressure initiates LPCI.

l.

Above trip setting, in conjunction with 1 ow reac tor water level, drywell high

pressure, 120 sec.

delay timer and CSS or RHR pump running, initiates ADS.

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 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.

2.

In only one trip system.

3.

Not considered in a trip system.

4.

Requires one channel from each physical location (there are 4 locations) in the steam line space.

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 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 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

NOTES FOR ABLE 2

B ont'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.

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.

Not required to be OPERABLE in the COLD SHUTDOWN CONDITION.

BFN Unit 1 3.2/4.2-24

4 CORE AND CONTA ME COOLI G SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.B Res dug Heat Removal S ste

~RfLRRS

. (LPCI and Containment Cooling) 4.5.B es dual Heat Remova S ste

~RHRS iLPCP and Containment Cooling)

8. If Specifications 3.5.B.1 through 3.5.B.7 are not met, an orderly shutdown shall be initiated and the reactor shall be placed in the COLD SHUTDOWN CONDITION within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

8.

No additional surveillance required.

9.

When the reactor vessel pressure is atmospheric and irradiated fuel is in the reactor vessel, at least one RHR loop with two pumps or two loops with one pump per loop shall be OPERABLE.

The pumps'ssociated diesel generators must also be OPERABLE.

9.

When the reactor vessel pressure is atmospheric, the RHR pumps and valves that are required to be OPERABLE shall be demonstrated to be OPERABLE per Specification 1.0.MM.

10. If the conditions of Specification 3.5.A.5 are met, LPCI and containment cooling are not required.

10.

No additional surveillance required.

ll.

When there is irradiated fuel in the reactor and the reactor is not in the COLD SHUTDOWN CONDITION, 2 RHR pumps and associated heat exchangers and valves on an adjacent unit must be OPERABLE and capable of supplying cross-connect capability except as specified in Specification 3.5.B.12 below.

(Note:

Because cross-connect capability is not a short-term requirement, a

component is not considered inoperable if cross-connect capability can be restored to service within 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.)

11.

The RHR pumps on the adjacent units which supply cross-connect capability shall be demonstrated to be OPERABLE per Specification 1.0.MM when the cross-connect capability is required.

BFN Unit 1 3.5/4.5-7

4 CORE A D CO COO I G SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.B Res dua eat Remova S stem

~RHRS (LPCI and Containment Cooling)

12. If one RHR pump or associated heat exchanger located on the unit cross-connection in the adjacent unit is inoperable for any reason (including valve inoperability, pipe break, etc.),

the reactor may remain in operation for a period not to exceed 30 days provided the remaining RHR pump and associated diesel generator are OPERABLE.

4.5.B Res dua eat Remova S ste

~RHRS (LPCI and Containment Cooling) 12.

No additional surveillance required.

13. If RHR cross-connection flow or heat removal capability is lost, the unit may remain in operation for a period not to exceed 10 days unless such capability is restored.

13.

No additional surveillance required.

14.

All recirculation pump discharge valves shall be OPERABLE PRIOR TO STARTUP (or closed if permitted elsewhere in these specifications).

14.

All recirculation pump discharge valves shall be tested for OPERABILITY during any period of COLD SHUTDOWN CONDITION exceeding 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, if OPERABILITY tests have not been performed during the preceding 31 days.

BFN Unit 1 3.5/4.5-8 AMENDMEN[f0. g G 9

4 CORE CO t

COOLI G SYSTE S

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.C RHR Service Water and E er enc, E ui ment Coolin Wate S stems EECWS Continued 4.5.C RHR Service Water and Emer enc E ui ment Coolin Water S stems EECWS Cont nued 4.

One of the Dl or D2 RHRSW pumps assigned to the RHR heat exchanger supplying the standby coolant supply connection may be inoperable for a period not to exceed 30 days provided the OPERABLE pump is aligned to supply the RHR heat exchanger header and the associated diesel generator and essential control valves are OPERABLE.

4.

No additional surveillance is required.

5.

The standby coolant supply capability may be inoperable for a period not to exceed 10 days.

6. If Specifications 3.5.C.2 through 3.5.C.5 are not
met, an orderly shutdown shall be initiated and the unit placed in the COLD SHUTDOWN CONDITION within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

7.

There shall be at least 2

RHRSW pumps, associated with the selected RHR pumps, aligned for RHR heat exchanger service for each reactor vessel containing irradiated fuel.

BFN Unit 1 3.5/4.5-12 AMEHOMENTg0. y g g

/4 5

CORE AND CONTAINME COOLING SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.D ui ment Area Coolers 4.5.D E ui ment ea Coolers 1.

The equipment area cooler associated with each RHR pump and the equipment area cooler associated with each set of core spray pumps (A and C

or B and D) must be OPERABLE at all times when the pump or pumps served by that specific cooler is considered to

,be OPERABLE.

l.

Each equipment area cooler is operated in conjunction with the equipment served by that particular cooler; therefore, the equipment area coolers are tested at the same frequency as the pumps which they serve.

2.

When an equipment area cooler is not OPERABLE, the pump(s) served by that cooler must be considered inoperable for Technical Specification purposes.

E.

Hi h Pressure Coolant In ection S stem HPC S

E.

H h Pressure Coolant In ection S stem HPCIS 1.

The HPCI system shall be OPERABLE whenever there is irradiated fuel in the reactor vessel and the reactor vessel pressure is greater than 150 psig, except in the COLD SHUTDOWN CONDITION or as specified in 3.5.E.2.

OPERABILITY shall be determined within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure reaches 150 psig from a COLD CONDITION, or alternatively PRIOR TO STARTUP by using an auxiliary steam supply.

a.

Simulated Automatic Actuation Test Once/18 months b.

Pump OPERA-BILITY Per Specification 1.0.MM c.

Motor Oper-ated Valve OPERABILITY Per Specification 1.0.MM 1.

HPCI Subsystem testing shall be performed as follows:

d.

Flow Rate at normal reactor vessel operating pressure Once/3 months BFN Unit 1 3.5/4.5-13

~

~

.5 4.

CORE A CO AINMENT COOLING SYSTE S

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.E Hi h Pressure Coo ant In ection 4.5.E Hi h Pressure Coolant In ection 4.5.E.1 (Cont'd) e.

Flow Rate at Once/18 150 psig months The HPCI pump shall deliver at least 5000 gpm during each flow rate test.

Once/Month f.

Verify that each valve (manual, power-

operated, or automatic) in the injection flow-path that is not locked, sealed, or otherwise secured in position, is in its correct* position.
2. If the HPCI system is inoperable, the reactor may remain in operation for a period not to exceed 7 days, provided the ADS,
CSS, RHRS (LPCI), and RCICS are OPERABLE.

2.

No additional surveillances are required.

3. If Specifications 3.5.E.l or 3.5.E.2 are not met, an orderly shutdown shall be initiated and the reactor vessel pressure shall be reduced to 150 psig or less within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Except that an automatic valve capable of automatic return to its ECCS position when an ECCS signal is present may be in a position for another mode of operation.

F.

Reactor Core Isolation Coolin F.

Reactor Core Isolation Coolin 1.

The RCICS shall be OPERABLE whenever there is irradiated fuel in the reactor vessel and the reactor vessel pressure is above 150 psig, except in the COLD SHUTDOWN CONDITION or as specified in 3.5.F.2.

OPERABILITY shall 1.

RCIC Subsystem testing shall be performed as follows:

a.

Simulated Auto-Once/18 matic Actuation months Test BFN Unit 1 3.5/4.5-14

3 4

CORE AND CO NME COOLI G SYS EMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.F Reacto Co e

so ation Coo in 4.5.F Reactor Core Iso at o

Coo i 3.5.F.l (Cont'd) be determined within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure reaches 150 psig from a COLD CONDITION or alternatively PRIOR TO STARTUP by using an auxiliary steam supply.

4.5.F.1 (Cont'd) b.

Pump OPERABILITY

c. Motor-Operated Valve OPERABILITY Per Specifi-cation 1.0.MM Per Specifi-cation 1.0.MM d.

Flow Rate at Once/3 normal reactor months vessel operating pressure e.

Flow Rate at 150 psig Once/18 months 2.

If the RCICS is inoperable, the reactor may remain in operation for a period not to exceed 7 days if the HPCIS is OPERABLE during such time.

3.

If Specifications 3.5.F.1 or 3.5.F.2 are not met, an orderly shutdown shall be initiated and the reactor shall be depressurized to less than 150 psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

The RCIC pump shall deliver at least 600 gpm during each flow test.

Once/Month f.

Verify that each valve (manual, power-

operated, or automatic) in the injection flowpath that is not locked,
sealed, or other-wise secured in position, is in its correct* position.

2.

No additional surveillances are required.

Except that an automatic valve capable of automatic return to its normal position when a signal is present may be in a position for another mode of operation.

BFN Unit 1 3.5/4.5-15 AMEHDMEI'tTNO. I7 3

4 CORE A

D CO AINME COOL G

S STEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.G Automatic De essur zation 4.5.G Automatic De ressurization Four of the six valves of the Automatic Depressurization System shall be OPERABLE:

1.

During each operating cycle the following tests shall be performed on the ADS:

(1) PRIOR TO STARTUP from a

COLD CONDITION, or, (2) whenever there is irradiated fuel in the reactor vessel and the reactor vessel pressure is greater than 105 psig, except in the COLD SHUT-DOWN CONDITION or as specified in 3.5.G.2 and 3.5.G.3 below.

a.

A simulated automatic actuation test shall be performed PRIOR TO STARTUP after each refueling outage.

Manual surveillance of the relief valves is covered in 4.6.D.2.

2. If three of the six ADS valves are known to be incapable of automatic operation, the reactor may remain in operation for a period not to exceed 7 days, provided the HPCI system is OPERABLE. (Note that the pressure relief function of these valves is assured by Section 3.6.D of these specifications and that this specification only applies to the ADS function.) If more than three of the six ADS valves are known to be incapable of automatic operation, an immediate orderly shutdown shall be initiated, with the reactor in a HOT SHUTDOWN CONDITION in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in a COLD SHUTDOWN CONDITION in the following 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br />.

2.

No additional surveillances are required.

3. If Specifications 3.5.G.1 and 3.5.G.2 cannot be met, an orderly shutdown will be BFN

~

Unit 1 3.5/4.5-16

5 4 CORE AND CO A

NT COOLI G

S S

EMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.G Automatic De ressurization 4.5.G Automatic De ressurizatio 3.5.G.3 (Cont'd) initiated and the reactor vessel pressure shall be reduced to 105 psig or less within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

H.

Maintenance o

F ed Dischar e

~Pi e

H.

Ma ntenance of Filled Dischar e

~Pi e

Whenever the core spray systems, LPCI, HPCI, or RCIC are required to be OPERABLE, the discharge piping from the pump discharge of these systems to the last block valve shall be filled.

The following surveillance requirements shall be adhered to assure that the discharge piping of the core spray

systems, LPCI, HPCI, and RCIC are filled:

The suction of the RCIC and HPCI pumps shall be aligned to the condensate storage

tank, and the pressure suppression chamber head tank shall normally be aligned to serve the discharge piping of the RHR and CS pumps.

The condensate head tank may be used to serve the RHR and CS discharge piping if the PSC head stank is unavailable.

The pressure indicators on the discharge of the RHR and CS pumps shall indicate not less than listed below.

1.

Every month and prior to the testing of the RHRS (LPCI and Containment Spray) and core spray system, the discharge piping of these systems shall be vented from the high point and water flow determined.

2.

Following any period where the LPCI or core spray systems have not been required to be OPERABLE, the discharge piping of the inoperable system shall be vented from the high point prior to the return of the system to service.

Pl-75-20 Pl-75-48 Pl-74-51 P1-74-65 48 psig 48 psig 48 psig 48 psig 3.

Whenever the HPCI or RCIC system is lined up to take suction from the condensate storage

tank, the discharge piping of the HPCI and RCIC shall be vented from the high point of the system and water flow observed on a monthly basis.

4.

When the RHRS and the CSS are required to be OPERABLE, the pressure indicators which monitor the discharge lines shall be monitored daily and the pressure recorded.

BFN Unit 1 3.5/4.5-17

4 PRIMARY SYSTE 0

ARY LIMITING CONDITIONS FOR OPERATION SURVEILLANCE RE UIREMENTS 4.6.C.

Coo ant eaka e

1.

a.

Any time irradiated fuel is in the reactor vessel and reactor coolant temperature is above 212 F, reactor coolant leakage into the primary containment from unidentified sources shall not exceed 5 gpm.

In addition, the total reactor coolant system leakage into the primary containment shall not exceed 25 gpm.

1.

Reactor coolant system leakage shall be checked by the sump and air sampling system and recorded at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

b.

Anytime the reactor is in RUN MODE, reactor coolant leakage into the primary containment from unidentified sources shall not increase by more than 2 gpm averaged over any 24-hour period in which the reactor is in the RUN MODE except as defined in 3.6.C.l.c below.

c.

During the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in the RUN MODE following

STARTUP, an increase in reactor coolant leakage into the primary containment of >2 gpm is acceptable as long as the requirements of 3.6.C.l.a are met.

BFN Unit 1 3.6/4.6-9 AMENDMENTNO. I3 7

4 PRIMARY SYSTE OUNDARY LIMITING CONDITIONS FOR OPERATION 3.6.C Coolant Leaka e

SURVEILLANCE REQUIREMENTS 4.6.C Coolant Leaka e

2.

Both the sump and air sampling systems shall be OPERABLE during REACTOR POWER OPERATION.

From and after the date that one of these systems is made or found to be inoperable for any reason, REACTOR POWER OPERATION is permissible only during the succeeding 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for the sump system or 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> for the air sampling system.

2.

With the air sampling system inoperable, grab samples shall be obtained and analyzed at least once every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

The air sampling system may be removed from service for a period of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for calibration, function testing, and maintenance without providing a temporary monitor.

3. If the condition in 1 or 2 above cannot be met, an orderly shutdown shall be initiated and the reactor shall be placed in the COLD SHUTDOWN CONDITION within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

D.

Relief Valves D.

Relief Valves 1.

When more than one relief valve is known to be failed, an orderly shutdown shall be initiated and the reactor depressurized to less than 105 psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

The relief valves are not required to be OPERABLE in the COLD SHUTDOWN CONDITION.

l.

Approximately one-half of all relief valves shall be bench-checked or replaced with a bench-checked valve each operating cycle.

All 13 valves will have been checked or replaced upon the completion of every second cycle.

2.

In accordance with Specification 1.0.MM, each relief valve shall be manually opened until thermocouples and acoustic monitors downstream of the valve indicate steam is flowing from the valve.

BFN Unit 1 3.6/4.6-10

3.6/4.6 BASES 3.6.C/4.6.C (Cont'd) suggest a reasonable margin of safety that such leakage magnitude would not result from a crack approaching the critical size for rapid propagation.

Leakage less than the magnitude specified can be detected reasonably in a matter of a few hours utilizing the available leakage detection

schemes, and if the origin cannot be determined in a reasonably short time, the unit should be shut down to allow further investigation and corrective action.

The two gpm limit for coolant leakage rate increase over any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period is a limit specified by the NRC (Reference 2).

This limit applies only during the RUN mode to avoid being penalized for the expected coolant leakage increase during pressurization.

The total leakage rate consists of all leakage, identified and unidentified, which flows to the drywell floor drain and equipment drain sumps

~

The capacity of the drywell floor sump pump is 50 gpm and the capacity of the drywell equipment sump pump is also 50 gpm.

Removal of 25 gpm from either of these sumps can be accomplished with considerable margin.

REFERENCE

1. Nuclear System Leakage Rate Limits (BFNP FSAR Subsection 4.10)
2. Safety Evaluation Report (SER) on IE Bulletin 82-03 D 4 D

Re ef Valves To meet the safety basis, 13 relief valves have been installed on the unit with a total capacity of 84.1 percent of nuclear boiler rated steam flow at a reference pressure of (1,105

+ 1 percent) psig.

The analysis of the worst overpressure transient, (3-second closure of all main steam line isolation valves) neglecting the direct scram (valve position scram) results in a maximum vessel pressure which, if a neutron flux scram is assumed considering 12 valves operable, results in adequate margin to the code allowable overpressure limit of 1,375 psig.

To meet operational

design, the analysis of the plant isolation transient (generator load reject with bypass valve failure to open) shows that 12 of the 13 relief valves limit peak system pressure to a value which is well below the allowed vessel overpressure of 1,375 psig.

'Experience in relief valve operation shows that a testing of 50 percent of the valves per year is adequate to detect failures or deteriorations.

The relief valves are benchtested every second operating cycle to ensure that their setpoints are within the g 1 percent tolerance.

The relief valves are tested in place in accordance with Specification 1.0.MN to establish that they will open and pass steam.

BFN Unit 1 3.6/4.6-30 AMENDMENT NO g~O

3.6/4.6

~BAS S

3.6.D/4.6.D (Cont'd)

The requirements established above apply when the nuclear system can be pressurized above ambient conditions.

These requirements are applicable at nuclear system pressures below normal operating pressures because abnormal operational transients could possibly start at these conditions such that eventual overpressure relief would be needed.

However, these transients are much less severe, in terms of pressure, than those starting at rated conditions.

The valves need not be functional when the vessel head is removed, since the nuclear system cannot be pressurized.

The relief valves are not required to be OPERABLE in the COLD SHUTDOWN CONDITION.

Overpressure protection is provided during hydrostatic tests by two of the relief valves whose relief setting has been established in conformance with ASME Section XI code requirements.

The capacity of one relief valve exceeds the charging capacity of the pressurization source used during hydrostatic testing.

Two relief valves are used to provide redundancy.

REFERENCES 1.

Nuclear System Pressure Relief System (BFNP FSAR Subsection 4.4) 2.

Amendment 22 in response to AEC Question 4.2 of December 6, 1971.

3.

"Protection Against Overpressure" (ASME Boiler and Pressure Vessel Code,Section III, Article 9) 4.

Browns Ferry Nuclear Plant Design Deficiency ReportTarget Rock Safety-Relief Valves, transmitted by J. E. Gilliland to F. E. Kruesi, August 29, 1973 5.

Generic Reload Fuel Application, Licensing Topical Report, NEDE-24011-P-A and Addenda 3.6.E/4.6.E

~Jet Pum s Failure of a jet pump nozzle assembly holddown mechanism, nozzle assembly and/or riser, would increase the cross-sectional flow area for blowdown following the design basis double-ended line break.

Also, failure of the diffuser would eliminate the capability to reflood the core to two-thirds height level following a recirculation line break.

Therefore, if a failure occurred, repairs must be made.

The detection technique is as follows.

With the two recirculation pumps balanced in speed to within + 5 percent, the flow rates in both recirculation loops will be verified by control room monitoring instruments.

If the two flow rate values do not differ by more than 10 percent, riser and nozzle assembly integrity has been verified.

BFN Unit 1 3.6/4.6-31

3.6/4.6 BASES 3.6.E/4.6.E (Cont'd)

If they do differ by 10 percent or more, the core flow rate measured by the jet pump diffuser differential pressure system must be checked against the core flow rate derived from the measured values of loop flow to core flow correlation. If the difference between measured and derived core flow rate is 10 percent or more (with the derived value higher) diffuser measurements will be taken to define the location within the vessel of failed jet pump nozzle (or riser) and the unit shut down for repairs.

If the potential blowdown flow area is increased, the system resistance to the recirculation pump is also reduced;

hence, the affected drive pump will "run out" to a substantially higher flow rate (approximately 115 percent to 120 percent for a single nozzle failure).

If the two loops are balanced in flow at the same pump speed, the resistance characteristics cannot have changed.

Any imbalance between drive loop flow rates would be indicated by the plant process instrumentation.

In addition, the affected jet pump would provide a

leakage path past the core thus reducing the core flow rate.

The reverse flow through the inactive jet pump would still be indicated by a positive differential pressure but the net effect would be a slight decrease (3 percent to 6 percent) in the total core flow measured.

This decrease, together with the loop flow increase, would result in a lack of correlation between measured and derived core flow rate.

Finally, the affected jet pump diffuser differential pressure signal would be reduced because the backflow would be less than the normal forward flow.

A nozzle-riser system failure could also generate the coincident failure of a jet pump diffuser body; however, the converse is not true.

The lack of any substantial stress in the jet pump diffuser body makes failure impossible without an initial nozzle-riser system failure.

3.6.F/4.6.F ecircu at on Pum 0 e at o

Steady-state operation without forced recirculation will not be permitted for more than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

And the start of a recirculation pump from the natural circulation condition will not be permitted unless the temperature difference between the loop to be started and the core coolant temperature is less than 75 F.

This reduces the positive reactivity insertion to an acceptably low value.

Requiring the discharge valve of the lower speed loop to remain closed until the speed of the faster pump is below 50% of its rated speed provides assurance when going from one-to-two pump operation that excessive vibration of the jet pump risers will not occur.

3.6.G/4.6.G Structural Inte rit The requirements for the reactor coolant systems inservice inspection program have been identified by evaluating the need for a sampling examination of areas of high stress and highest probability of failure in the system and the need to meet as closely as possible the requirements of Section XI, of the ASME Boiler and Pressure Vessel Code.

BFN Unit 3.6/4.6-32

3.6/4.6 BASES 3.6.G/4.6.G (Cont'd)

The program reflects the built-in limitations of access to the reactor coolant systems.

It is intended that the required examinations and inspection be completed during each 10-year interval.

The periodic examinations are to be done during refueling outages or other extended plant shutdown periods.

Only proven nondestructive testing techniques will be used.

More frequent inspections shall be performed on certain circumferential pipe welds as listed in Section 4.6.G.4 to provide additional protection against pipe whip.

These welds were selected in respect to their distance from hangers or supports wherein a failure of the weld would permit the unsupported segments of pipe to strike the drywell wall or nearby auxiliary systems or control systems.

Selection was based on judgment from actual plant observation of hanger and support locations and review of drawings.

Inspection of all these welds during each 10-year inspection interval will result in three additional examinations above the requirements of Section XI of ASME Code.

An augmented inservice surveillance program is required to determine whether any stress corrosion has occurred in any stainless steel piping, stainless components, and highly-stressed alloy steel such as hanger

springs, as a result of environmental conditions associated with the March 22, 1975 fire.

g~EF RELICS 1.

Inservice Inspection and Testing (BFNP FSAR Subsection 4.12) 2.

Inservice Inspection of Nuclear Reactor Coolant Systems,Section XI, ASME Boiler and Pressure Vessel Code 3.

ASME Boiler and Pressure Vessel Code,Section III (1968 Edition) 4.

American Society for Nondestructive Tgsting No. SNT-TC-1A (1968 Edition) 5.

Mechanical Maintenance Instruction 46 (Mechanical Equipment,

Concrete, and Structural Steel Cleaning Procedure for Residue From Plant Fire Units 1 and 2) 6.

Mechanical Maintenance Instruction 53 (Evaluation of Corrosion Damage of Piping Components Which Were Exposed to Residue From March 22, 1975 Fire) 7.

Plant Safety Analysis (BFNP FSAR Subsection 4.12)

BFN Unit 1 3.6/4.6-33

UNIT 2 EFFECTIVE PAGE LIST REMOVE INSERT 3.2/4.2-14 3.2/4.2-15 3.2/4.2-16 3.2/4.2-17 3.2/4.2-23 3.2/4.2-24 3.5/4.5-7 3.5/4.5-8 3.5/4.5-12 3.5/4.5-13 3.5/4.5-14 3.5/4.5-15 3.5/4.5-16 3.5/4.5-17 3.6/4.6-9 3.6/4.6-10 3.6/4.6-30 3.6/4.6-31 3.6/4.6-32.

3.6/4.6-33 3.2/4.2-14 3.2/4.2-15 3.2/4.2-16*

3.2/4.2-17 3.2/4.2-23+

3.2/4.2-24 3.5/4.5-7 3.5/4.5-8*

3.5/4.5-12*

3.5/4.5-13 3.5/4.5-14 3.5/4.5-15*

3.5/4.5-16 3.5/4.5-17*

3.6/4.6-9+

3.6/4.6-10 3.6/4.6-30*

3.6/4.6-31 3.6/4.6-32*

3.6/4.6-33*

  • Denotes overleaf or spillover page.

TABLE 3.2.B INSTRUMENTATION THAT INITIATES OR CONTROLS THE CORE AND CONTAINMENT COOLING SYSTEMS Minimum No.

Operable Per

~Tri '~l Fun tion Tri L vel e

in Action R mark 2

2(16)

Instrument Channel Reactor Low Water Level (LIS-3-58A-D)

Instrument Channel Reactor Low Water Level (LIS-3-58A-D)

Instrument Channel Reactor Low Water Level (LS-3-58A-0)

Instrument Channel-Reactor Low Water Level (LS-3-58A-D)

> 470u above vessel zero.

A

> 470" above vessel zero.

A

> 378" above vessel zero.

A

> 378" above vessel zero.

A 1.

Below trip setting initiates HPCI.

1.

Multiplier relays initiate RCIC.

1.

Below trip setting initiates CSS.

Hultiplier relays initiate LPCI.

2.

Multiplier relay from CSS initiates accident signal (15).

l.

Below trip settings, in conjunction with drywell, high pressure.

low water level permissive, 105 sec.

delay timer and CSS or RHR pump running, initiates ADS.

2.

Below trip settings, in conjunction with low reactor water level permissive, 105 sec.

delay timer, 12 1/2 min. delay timer, CSS or RHR pump running, initiates ADS.

1(16)

Instrument Channel Reactor Low Water Level Permissive (LIS-3-184, 185)

Instrument Channel-Reactor Low Water Level (LIS-3-52 and LIS-3-62A)

> 544" above vessel zero.

A

> 312 5/16" above vessel zero.

A (2/3 core height) 1.

Below trip setting permissive for initiating signals on ADS.

1.

Below trip setting prevents inadvertent operation of containment spray during accident condition.

The automatic initiation capability of this instrument channel is not required to be OPERABLE while the Reactor Vessel water level monitoring modification is being performed.

Hanual initiation capability of the associated system will be available during that time the automatic initiation logic is out-of-service.

~

~

TABLE 3.2.8 (Continued)

Hinimum No.

Operable Per T~ri S

i 2(18) 2(18)

Func i n Instrument Channel-Drywell High Pressure (PIS-64-58 E-H)

Instrument Channel-Drywell High Pressure (PIS-64-58 A-D)

Tri L v 1

1< p<2.5 psig

< 2.5 psig in

~AinnA'emarks l.

Below trip setting prevents inadvertent operation of containment spray during accident conditions.

l.

Above trip setting in con-junction with low reactor pressure initiates CSS.

Hultiplier relays initiate HPCI.

2(18) 2(16) (18)

Instrument Channel-Drywell High Pressure (PIS-64-58A-D)

Instrument Channel-Drywell High Pressure (PIS-64-57A-D)

< 2.5 psig

< 2.5 psig 2.

Hultiplier relay from CSS initiates ace ident s ignal. (15) l.

Above trip setting in conjunction with low reactor pressure initiates LPCI.

1.

Above trip setting, in conjunction with low reactor water level, low reactor water level permissive, 105 sec.

delay timer and CSS or RHR pump running, initiates ADS.

~

~

TABLE 3.2.B (Continued)

Hinimum No.

Operable Per

~Tri S

1 Fun ti n Instrument Channel Reactor Low Pressure (PIS-3-74 A 8 B)

(PIS-68-95, 96)

Instrument Channel-Reactor Low Pressure (PS-3-74 A 5 B)

(PS-68-95, 96)

Instrument Channel Reactor Low Pressure (PS-68-93 8 94, SW 01)

Tri L vel Set in 450 psig + 15 230 psig + 15 100 psig + 15 A~ion Remarks 1.

Below trip setting permissive for opening CSS and LPCI admission valves.

1.

Recirculation discharge valve actuation.

l.

Below trip setting in conjunction with containment isolation signal and both suction valves open will close RHR (LPCI) admission valves.

Core Spray Auto Sequencing 6< t <8 sec.

Timers (5) 1.

With diesel power 2.

One per motor LPCI Auto Sequencing Timers (5) 0< t <1 sec.

1.

With diesel power 2.

One per motor RHRSW Al, B3, Cl, and 03 13< t <15 sec.

Timers 1.

With diesel power 2.

One per pump Core Spray and LPCI Auto Sequencing Timers (6)

RHRSW Al, B3, Cl, and 03 Timers 0< t <1 sec.

6< t <8 sec.

12< t <16 sec.

18< t <24 sec.

27< t < 29 sec.

1.

With normal power 2.

One per CSS motor 3.

Two per RHR motor 1.

With normal power 2.

One per pump

TABLE 3.2.B (Continued)

Minimum No.

Operable Per Iri~!Lbll 1(16)

Function AOS Timer Tri L v 1

tin 105 sec

+ 7

~AI II R marks 1.

Above trip setting in conjunction with low reactor water level permissive, low reactor water level, high drywell pressure or high drywell pressure bypass timer timed out, and RHR or CSS pumps running, initiates ADS.

1(16) 1(3)

Instrument Channel-RHR Discharge Pressure 100 +10 psig Instrument Channel 185 +10 psig CSS Pump Discharge Pressure Core Spray Sparger to 2 psid +0.4 Reactor Pressure Vessel d/p RHR (LPCI) Trip System bus N/A power monitor Core Spray Trip System bus N/A power monitor ADS Trip System bus power N/A monitor ADS Timer (12 1/2 min.)

12 1/2 min. + 2 (High Drywell Pressure Bypass Timer) l.

Above trip setting, in conjunction with low reactor water level permissive, low reactor water level, 105 sec.

delay timer, and RHR or CSS pumps running, initiates AOS.

l.

Below trip setting defers ADS actuation.

l.

Below trip setting defers AOS actuation.

Alarm to detect core sparger pipe break.

1.

Monitors availability of power to logic systems.

1.

Monitors avail abil i ty of power to logic systems.

1.

Monitors availability of power to logic systems and valves.

I~

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 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.

2.

In only one trip system.

3.

Not considered in a trip system.

4.

Requires one channel from each physical location (there are 4 locations) in the steam line space.

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 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 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 2 3.2/4.2-23

NOTES OR TABLE 2

B Cont'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.

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.

Not required to be OPERABLE in the COLD SHUTDOWN CONDITION.

BFN Unit 2 3.2/4.2-24

4.

CORE A D CO AINMENT COOLI G SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.B Residual Heat Remova S ste

~RHRS (LPCI and Containment Cooling) 4.5.B Res dua Heat Remova S stem

~RHRS (LPCI and Containment Cooling)

8. If Specifications 3.5.B.l through 3.5.B.7 are not met, an orderly shutdown shall be initiated and the reactor shall be placed in the COLD SHUTDOWN CONDITION within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

8.

No additional surveillance required.

9.

When the reactor vessel pressure is atmospheric and irradiated fuel is in the reactor vessel, at least one RHR loop with two pumps or two loops with one pump per loop shall be OPERABLE.

The pumps'ssociated diesel generators must also be OPERABLE.

9.

When the reactor vessel pressure is atmospheric, the RHR pumps and valves that are required to be OPERABLE shall be demonstrated to be OPERABLE per Specification 1.0.MM.

10. If the conditions of Specification 3.5.A.5 are met, LPCI and containment cooling are not required.

10.

No additional surveillance required.

ll.

When there is irradiated fuel in the reactor and the reactor is not in the COLD SHUTDOWN CONDITION, 2 RHR pumps and associated heat exchangers and valves on an adjacent unit must be OPERABLE and capable of supplying cross-connect capability except as specified in Specification 3.5.B.12 below.

(Note:

Because cross-connect capability is not a short-term requirement, a component is not considered inoperable if cross-connect capability can be restored to service within 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.)

11.

The RHR pumps on the adjacent units which supply cross-connect capability shall be demonstrated to be OPERABLE per Specification 1.0.MM when the cross-connect capability is required.

BFN Unit 2 3.5/4.5-7

4 CORE AND CO COOLING SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.B esidual Heat e oval S ste

~RHRS (LPCI and Containment Cooling)

12. If three RHR pumps or associated heat exchangers located on the unit cross-connection in the adjacent units are inoperable for any reason (including valve inoperability, pipe break, etc.),

the reactor may remain in operation for a period not to exceed 30 days provided the remaining RHR pump and associated diesel generator are OPERABLE.

13. If RHR cross-connection flow or heat removal capability is lost, the unit may remain in operation for a period not to exceed 10 days unless such capability is restored.

4.5.B es dua eat Removal S stem PHRS (LPCI and Containment Cooling) 12.

No additional surveillance required.

13.

No additional surveillance required.

14.

All recirculation pump discharge valves shall be OPERABLE PRIOR TO STARTUP (or closed if permitted elsewhere in these specifications).

14.

All recirculation pump discharge valves shall be tested for OPERABILITY during any period of COLD SHUTDOWN CONDITION exceeding 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, if OPERABILITY tests have not been performed during the preceding 31 days.

BFN Unit 2 3.5/4.5-8 AMENDMENTNO. X 6 9

4 CORE AND CO COOLI G SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.C RHR Service Water and Emer enc E ui ment Coo in Wate S stems EECWS Cont nued 4.5.C RHR Service Water and Emer enc u

ment, Coolin Water S stems EECWS Continued 4.

Three of the Dl, D2, Bl, B2 RHRSW pumps assigned to the RHR heat exchanger supplying the standby coolant supply connection may be inoperable for a period not to exceed 30 days provided the OPERABLE pump is aligned to supply the RHR heat exchanger header and the associated diesel generator and essential control valves are OPERABLE.

4.

No additional surveillance is required.

5.

The standby coolant supply capability may be inoperable for a period not to exceed 10 days.

6. If Specifications 3.5.C.2 through 3.5.C.5 are not
met, an orderly shutdown shall be initiated and the unit placed in the COLD SHUTDOWN CONDITION within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

7.

There shall be at least 2

RHRSW pumps, associated with the selected RHR pumps, aligned for RHR heat exchanger service for each reactor vessel containing irradiated fuel.

BFN Unit 2 3.5/4.5-12 AMfNDMEHTNO. I 6 9

5 CORE CO COOLI G

S STEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.D ent A ea Coo ers 4.5.D ui ment Area Coole s 1.

The equipment area cooler associated with each RHR pump and the equipment area cooler associated with each set of core spray pumps (A and C

or B and D) must be OPERABLE at all times when the pump or pumps served by that specific cooler is considered to be OPERABLE.

l.

Each equipment area cooler is operated in conjunction with the equipment served by that particular cooler; therefore, the equipment area coolers are tested at the same frequency as the pumps which they serve.

2.

When an equipment area cooler is not OPERABLE, the pump(s) served by that cooler must be considered inoperable for Technical Specification purposes.

E.

Hi h Pressure Coolant In ect o

E.

Hi h Pressure Coolant In 'ection S stem HPCIS The HPCI system shall be OPERABLE whenever there is irradiated fuel in the reactor vessel and the reactor vessel pressure is greater than 150 psig, except in the COLD SHUTDOWN CONDITION or as specified in 3.5.E.2.

OPERABILITY shall be determined within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure reaches 150 psig from a COLD CONDITION, or alter-natively PRIOR TO STARTUP by using an auxiliary steam supply.

a.

Simulated Automatic Actuation Test Once/18 months b.

Pump OPERA-BILITY Per Specification 1.0.MM c.

Motor Oper-ated Valve OPERABILITY Per Specification 1.0.MM 1.

HPCI Subsystem testing shall be performed as follows:

d.

Flow Rate at normal reactor vessel operating pressure Once/3 months BFN Unit 2 3.5/4.5-13

4.5 CORE AND CO AINME COOLING SYS E

S LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.E Hi Pressure Coo ant In ection 4.5.E Hi h Pressure Coolant In ection 4.5.E.l (Cont'd) e.

Flow Rate at Once/18 150 psig months The HPCI pump shall deliver at least 5000 gpm during each flow rate test.

Once/Month f.

Verify that each valve (manual, power-

operated, or automatic) in the injection flow-path that is not locked, sealed, or otherwise secured in position, is in its correct* position.
2. If the HPCI system is inoperable, the reactor may remain in operation for a period not to exceed 7 days, provided the ADS,
CSS, RHRS(LPCI),

and RCICS are OPERABLE.

2.

No additional surveillances are required.

3. If Specifications 3.5.E.l or 3.5.E.2 are not met, an orderly shutdown shall be initiated and the reactor'essel pressure shall be reduced to 150 psig or less within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Except that an automatic valve capable of automatic return to its ECCS position when an ECCS signal is present may be in a position for another mode of operation.

F.

Reactor Core Isolation Coolin F.

Reactor Core Isolation Coolin 1.

The RCICS shall be OPERABLE whenever there is irradiated fuel in the reactor vessel and the reactor vessel pressure is above 150 psig, except in the COLD SHUTDOWN CONDITION or as specified in 3.5.F.2.

OPERABILITY shall 1.

RCIC Subsystem testing shall be performed as follows:

a.

Simulated Auto-Once/18 matic Actuation months Test BFN Unit 2 3.5/4.5-14

4 CORE A

CO COOLI G SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.F.

Reactor Co e Iso ation Coolin 4.5.F Reactor Core Isolation Coolin 3.5.F.l (Cont'd) be determined within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure reaches 150 psig from a COLD CONDITION or alternatively PRIOR TO STARTUP by using an auxiliary steam supply..

4.5.F.1 (Cont'd) b.

Pump OPERABILITY

c. Motor-Operated Valve OPERABILITY Per Specifi-cation 1.0.MM Per Specifi-cation 1.0.MM d.

Flow Rate at Once/3 normal reactor months vessel operating pressure e.

Flow Rate at Once/18 150 psig months The RCIC pump shall deliver at least 600 gpm during each flow test.

2. If the RCICS is inoperable, the reactor may remain in operation for a period not to exceed 7 days if the HPCIS is OPERABLE during such time.
3. If Specifications 3.5.F.1 or 3.5.F.2 are not met, an orderly shutdown shall be initiated and the reactor shall be depressurized to less than 150 psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Once/Month f.

Verify that each valve (manual, power-

operated, or automatic) in the injection flowpath that is not locked,
sealed, or other-wise secured in position, is in its correct* position.

2.

No additional surveillances are required.

Except that an automatic valve capable of automatic return to its normal position when a signal is present may be in a position for another mode of operation.

BFN Unit 2 3.5/4. 5-15 AMENOMBlTWO. X V 6

4 CORE AND CONTA NMENT COOLING SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.G Automat c De ressu atio 4.5.G utomatic De ressurizatio 1.

Four of the six valves of the Automatic Depressurization System shall be OPERABLE:

1.

During each operating cycle the following tests shall be performed on the ADS:

(1) PRIOR TO STARTUP from a

COLD CONDITION, or, (2).whenever there is irradiated fuel in the reactor vessel and the reactor vessel pressure is greater than 105 psig, except in the COLD SHUT-DOWN CONDITION or as specified in 3.5.G.2 and 3.5.G.3 below.

a.

A simulated automatic actuation test shall be performed PRIOR TO STARTUP after each refueling outage.

Manual surveillance of the relief valves is covered in 4.6.D.2.

2. If three of the six ADS valves are known to be incapable of automatic operation, the reactor may remain in operation for a period not to exceed 7.days, provided the HPCI system is OPERABLE. (Note that the pressure relief function of these valves is assured by Section 3.6.D of these specifications and that this specification only applies to the ADS function.) If more than three of the six ADS valves are known to be incapable of automatic operation, an immediate orderly shutdown shall be initiated, with the reactor in a HOT SHUTDOWN CONDITION in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in a COLD SHUTDOWN CONDITION in the following 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br />.

2.

No additional surveillances are required.

3. If Specifications 3.5.G.l and 3.5.G.2 cannot be met, an orderly shutdown will be initiated and the reactor BFN Unit 2 3.5/4.5-16

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.G Automatic De ressur zat o

4.5.G Automatic De ressurization 3.5.G.3 (Cont'd) vessel pressure shall be reduced to 105 psig or less within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

H.

Maintenance of Filled Dischar e

~Pi e

H.

aintenance of F lied Dischar e

~Pi e

Whenever the core spray systems, LPCI, HPCI, or RCIC are required to be OPERABLE, the discharge piping from the pump discharge of these systems to the last block valve shall be filled.

The following surveillance requirements shall be adhered to assure that the discharge piping of the core spray

systems, LPCI, HPCI, and RCIC are filled:

The suction of the RCIC and HPCI pumps shall be aligned to the condensate storage

tank, and the pressure suppression chamber head tank shall normally be aligned to serve the discharge piping of the RHR and CS pumps.

The condensate head tank may be used to serve the RHR and CS discharge piping if the PSC head tank is unavailable.

The pressure indicators on the discharge of the RHR and CS pumps shall indicate not less than listed below.

l.

Every month and prior to the testing of the RHRS (LPCI and Containment Spray) and core spray system, the discharge piping of,,these systems shall be vented from the high point and water flow determined.

2.

Following any period where the LPCI or core spray systems have not been required to be OPERABLE, the discharge piping of the inoperable'ystem shall be vented from the high point prior to the return of the system to service.

Pl-75-20 Pl-75-48 Pl-74-51 Pl-74-65 48 psig 48 psig 48 psig 48 psig 3.

Whenever the HPCI or RCIC system is lined up to take suction from the condensate storage

tank, the discharge piping of the HPCI and RCIC shall be vented from the high point of the system and water flow observed on a monthly basis.

4.

When the RHRS and the CSS are required to be OPERABLE, the pressure indicators which monitor the discharge lines shall be monitored daily and the pressure recorded.

BFN Unit 2 3.5/4.5-17

4 PRIMAR S

S OUNDAR LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.6.C.

Coolant Leaka e

4.6.C.

Coolant Leaka e

lo a ~

Any time irradiated fuel is in the

~ reactor vessel and reactor coolant temperature is above 212'F, reactor coolant leakage into the primary containment from unidentified sources shall not exceed 5 gpm.

In addition, the total reactor coolant system leakage into the primary containment shall not exceed 25 gpm.

1.

Reactor coolant system leakage shall be checked by the sump and air sampling system and recorded at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

b.

Anytime the reactor is in RUN mode, reactor coolant leakage into the primary containment from unidentified sources shall not increase by more than 2 gpm averaged over any 24-hour period in which the reactor is in the RUN mode except as defined in 3.6.C.l.c below.

c ~

During. the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in the RUN mode following

STARTUP, an increase in reactor coolant leakage into the primary containment of >2 gpm is acceptable as long as the requirements of 3.6.C.l.a are met.

BFN Unit 2 3.6/4.6-9 AMENDMEHTtl0. I~ 3

4 PRIMARY SYSTE OUNDARY LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 4.6.C Coo ant Leaka e

2.

Both the sump and air sampling systems shall be OPERABLE during REACTOR POWER OPERATION.

From and after the date that one of these systems is made or found to be inoperable for any reason, REACTOR POWER OPERATION is permissible only during the succeeding 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for the sump system or 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> for the air sampling system.

2.

With the air sampling system inoperable, grab samples shall be obtained and analyzed at least once every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

The air sampling system may be removed from service for a period of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for calibration, function testing, and maintenance without providing a temporary monitor.

3. If the condition in 1 or 2 above cannot be met, an orderly shutdown shall be initiated and the reactor shall be placed in the COLD SHUTDOWN CONDITION within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

D.

Re ef Valves 1.

When more than one relief valve is known to be failed, an orderly shutdown shall be initiated and the reactor depressurized to less than 105 psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

The relief valves are not required to be OPERABLE in the COLD SHUTDOWN CONDITION.

l.

Approximately one-half of all relief valves shall be bench-checked or replaced with a bench-checked valve each operating cycle.

All 13 valves will have been checked or replaced upon the completion of every second cycle.

2.

In accordance with Specification 1.0.MM, each relief valve shall be manually opened until thermocouples and acoustic monitors downstream of the valve indicate steam is flowing from the valve.

BFN Unit 2 3.6/4.6-10

s.s/e.s

~ssEs 0

3.6.B/4.6.C (Cont'd) five gpm, as specified in 3.6.C, the experimental and analytical data suggest a reasonable margin of safety that such leakage magnitude would not result from a crack approaching the critical size for rapid propagation.

Leakage less than the magnitude specified can be detected reasonably in a matter of a few hours utilizing the available leakage detection

schemes, and if the origin cannot be determined in a reasonably short time, the unit should be shut down to allow further investigation and corrective action.

The 2 gpm limit for coolant leakage rate increases over any 24-hour period is a limit specified by the NRC (Reference 2).

This limit applies only during the RUN mode to avoid being penalized for the expected coolant leakage increase during pressurization.

The total leakage rate consists of all leakage, identified and unidentified, which flows to the drywell floor drain and equipment drain sumps.

The capacity of the drywell floor sump pump is 50 gpm and the capacity of the drywell equipment sump pump is also 50 gpm.

Removal of 25 gpm from either of these sumps can be accomplished with considerable margin.

REFERENCE

1. Nuclear System Leakage Rate Limits (BFNP FSAR Subsection 4.10)
2. Safety Evaluation Report (SER) on IE Bulletin 82-03 3.6.D/4.6.D Relief Valves To meet the safety basis, 13 relief valves have been installed on the unit with a total capacity of 84.1 percent of nuclear boiler rated steam flow.

The analysis of the worst overpressure transient, (3-second closure of all main steam line isolation valves) neglecting the direct scram (valve position scram) results in a maximum vessel pressure

which, if a neutron flux scram is assumed considering 12 valves
OPERABLE, results in adequate margin to the code allowable overpressure limit of 1,375 psig.

To meet operational design, the analysis of the plant isolation transient (generator load reject with bypass valve failure to open) shows that 12 of the 13 relief valves limit peak system pressure to a

value which is well below the allowed vessel overpressure of 1,375 psig.

Experience in relief valve operation shows that a testing of 50 percent of the valves per year is adequate to detect failures or deteriorations.

The relief valves are benchtested every second operating cycle to ensure that their setpoints are within the

+

1 percent tolerance.

The relief valves are tested in place in accordance with Specification 1.0.MM to establish that they will open and pass steam.

BFN Unit 2 3.6/4.6-30 AMENOMENT50. I P g

3.6/4.6 BASES 3.6.D/4.6.D (Cont'd)

The requirements established above apply when the nuclear system can be pressurized above ambient conditions.

These requirements are applicable at nuclear system pressures below normal operating pressures because abnormal operational transients could possibly start at these conditions such that eventual overpressure relief would be needed.

However, these transients are much less severe, in terms of pressure, than those starting at rated conditions.

The valves need not be functional when the vessel head is removed, since the nuclear system cannot be pressurized.

The relief valves are not required to be OPERABLE in the COLD SHUTDOWN CONDITION.

Overpressure protection is provided during hydrostatic tests by two of the relief valves whose relief setting has been established in conformance with ASME Section ZI code requirements.

The capacity of one relief valve exceeds the charging capacity of the pressurization source used during hydrostatic testing.

Two relief valves are used to provide redundancy.

~REPERE RES 1.

Nuclear System Pressure Relief System (BFNP FSAR Subsection 4.4) 2.

Amendment 22 in response to AEC Question 4.2 of December 6, 1971.

3.

"Protection Against Overpressure" (ASME Boiler and Pressure Vessel Code,Section III, Article 9) 4.

Browns Ferry Nuclear Plant Design Deficiency ReportTarget Rock Safety-Relief Valves, transmitted by J. E. Gilleland to F. E. Kruesi, August 29, 1973 5.

Generic Reload Fuel Application, Licensing Topical Report, NEDE-24011-P-A and Addenda 3.6.E/6.6.E J~et Pum s Failure of a jet pump nozzle assembly holddown mechanism, nozzle assembly and/or riser, would increase the cross-sectional flow area for blowdown following the design basis double-ended line break.

Also, failure of the diffuser would eliminate the capability to reflood the core to two-thirds height level following a recirculation line break.

Therefore, if a failure occurred, repairs must be made.

The detection technique is as follows.

With the two recirculation pumps balanced in speed to within g 5 percent, the flow rates in both recirculation loops will be verified by control room monitoring instruments.

If the two flow rate values do not differ by more than 10 percent, riser and nozzle assembly integrity has been verified.

BFN Unit 2 3.6/4.6-31

3.6/4.6 Q~S S

3.6.E/4.6.E (Cont'd)

If they do differ by 10 percent or more, the core flow rate measured by the jet pump diffuser differential pressure system must be checked against the core flow rate derived from the measured values of loop flow to core flow correlation. If the difference between measured and derived core flow rate is 10 percent or more (with the derived value higher) diffuser measurements will be taken to define the location within the vessel of failed jet pump nozzle (or riser) and the unit shut down for repairs.

If the potential blowdown flow area is increased, the system resistance to the recirculation pump is also reduced;

hence, the affected drive pump will "run out" to a substantially higher flow rate (approximately 115 percent to 120 percent for a single nozzle failure).

If the two loops are balanced in flow at the same pump speed, the resistance characteristics cannot have changed.

Any imbalance between drive loop flow rates would be indicated by the plant process instrumentation.

In addition, the affected jet pump would provide a leakage path past the core thus reducing the core flow rate.

The reverse flow through the inactive jet pump would still be indicated by a positive differential pressure but the net effect would be a slight decrease (3 percent to 6 percent) in the total core flow measured.

This decrease, together with the loop flow increase, would result in a lack of correlation between measured and derived core flow rate.

Finally, the affected jet pump diffuser differential pressure signal would be reduced because the backflow would be less than the normal forward flow.

A nozzle-riser system failure could also generate the coincident failure of a jet pump diffuser body; however, the converse is not true.

The lack of any substantial stress in the jet pump diffuser body makes failure impossible without an initial nozzle-riser system failure.

3.6.F/4.6.F Recirculation Pum 0 erat on Operation without forced recirculation is permitted for up to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is not in the RUN mode.

And the start of a recirculation pump from the natural circulation condition will not be permitted unless the temperature difference between the loop to be started and the core coolant temperature is less than 75 F.

This reduces the positive reactivity insertion to an acceptably low value.

Requiring at least one recirculation pump to be operable while in the RUN mode provides p'rotection against the potential occurrence of core thermal-hydraulic instabilities at low flow conditions.

Requiring the discharge valve of the lower speed loop to remain closed until the speed of the faster pump is below 50% of its rated speed provides assurance when going from one-to-two pump operation that excessive vibration of the jet pump risers will not occur.

BFN Unit 2 3.6/4.6-32

3.6/4.6 BASES 3.6.G/4.6.G St uctu al Inte rit The requirements for the reactor coolant systems inservice inspection program have been identified by evaluating the need for a sampling examination of areas of high stress and highest probability of failure in the system and the need to meet as closely as possible the requirements of Section XI, of the ASME Boiler and Pressure Vessel Code.

The program reflects the built-in limitations of access to the reactor coolant systems.

It is intended that the required examinations and inspection be completed during each 10-year interval.

The periodic examinations are to be done during refueling outages or other extended plant shutdown periods.

Only proven nondestructive testing techniques will be used.

More frequent inspections shall be performed on certain circumferential pipe welds as listed in Section 4.6.G.4 to provide additional protection against pipe whip.

These welds were selected in respect to their distance from hangers or supports wherein a failure of the weld would permit the unsupported segments of pipe to strike the drywell wall or nearby auxiliary systems or control systems.

Selection was based on judgment from actual plant observation of hanger and support locations and review of drawings.

Inspection of all these welds during each 10-year inspection interval will result in three additional examinations above the requirements of Section XI of ASME Code.

An augmented inservice surveillance program is required to determine whether any stress corrosion has occurred in any stainless steel piping, stainless components, and highly-stressed alloy steel such as hanger

springs, as a result of environmental conditions associated with the March 22, 1975 fire.

REFERENCES 1.

Inservice Inspection and Testing (BFNP FSAR Subsection 4.12) 2.

Inservice Inspection of Nuclear Reactor Coolant Systems,Section XI, ASME Boiler and Pressure Vessel Code 3.

ASME Boiler and Pressure Vessel Code,Section III (1968 Edition) 4.

American Society for Nondestructive Testing No. SNT-TC-1A (1968 Edition) 5.

Mechanical Maintenance Instruction 46 (Mechanical Equipment,

Concrete, and Structural Steel Cleaning Procedure for Residue From Plant Fire Units 1 and 2) 6.

Mechanical Maintenance Instruction 53 (Evaluation of Corrosion Damage of Piping Components Which Were Exposed to Residue From March 22, 1975 Fire) 7.

Plant Safety Analysis (BFNP FSAR Subsection 4.12)

BFN Unit 2 3.6/4.6-33

UNIT 3 EFFECTIVE PAGE LIST REMOVE INSERT 3.2/4.2-14 3.2/4.2-15 3.2/4.2-22 3.2/4.2-23 3.5/4.5-7 3.5/4.5-8 3.5/4.5-12 3.5/4.5-13 3.5/4.5-14 3.5/4.5-15 3.5/4.5-16 3.5/4.5-17 3.6/4.6-9 3.6/4.6-10 3.6/4.6-30 3.6/4.6-31 3.6/4.6-32 3.6/4.6-33 3.2/4.2-14 3.2/4.2-15 3.2/4.2-22*

3.2/4.2-23 3.5/4.5-7 3.5/4.5-8*

3.5/4.5-12*

3.5/4.5-13 3.5/4.5-14 3.5/4.5-15*

3.5/4.5-16 3.5/4.5-17*

3.6/4.6-9*

3.6/4.6-10 3.6/4.6-30*

3.6/4.6-31 3.6/4.6-32*

3.6/4.6-33*

+Denotes overleaf or spillover page.

TABLE 3.2.8 INSTRUMENTATION THAT INITIATES OR CONTROLS THE CORE AND CONTAINHENT COOLING SYSTEHS Hinimum No.

Operable Per

~lri Sr 1 Func ion Instrument Channel-Reactor Low Water Level Instrument Channel Reactor Low Mater Level Tri Level ettin

> 470" above vessel zero.

> 470" above vessel zero.

Action Remarks 1.

Below trip setting initiates HPCI 1.

Multiplier relays initiate RCIC.

Instrument Channel-Reactor Low Water Level (LIS-3-58A-D, SW01)

> 378" above vessel zero.

A 1.

Below trip setting initiates CSS.

Hultiplier relays initiate LPCI.

2.

Hultiplier relay from CSS initiates accident signal (15).

2(16)

Instrument Channel-

> 378" above vessel zero.

A Reactor Low Mater Level (LIS-3-58A-D, SW42) 1.

Below trip settings, in conjunction with drywell high pressure, low water level permissive, 120 sec.

delay timer and CSS or RHR pump running, initiates ADS.

1(16)

Instrument Channel Reactor Low Mater Level Permissive (LIS-3-184 4

185, SM41)

Instrument Channel Reactor Low Water Level (LITS-3-52 and 62, SW41)

> 544" above vessel zero.

A

> 312 5/16" above vessel zero.

A (2/3 core height) 1.

Below trip setting permissive for initiating signals on ADS.

1.

Below trip setting prevents inadvertent operation of containment spray during accident condition.

TABLE 3.2.B (Continued)

Hinimum No.

Operable Per

~Tri >~~1 2(18) 2(18) 2(18) 2(16) (18)

Func i n Instrument Channel-Drywell High Pressure (PS-64-58 E-H)

Instrument Channel-Orywell High Pressure (PS-64-58 A-D, SW¹2)

Instrument Channel-Drywell High Pressure (PS-64-58A-O, SW¹1)

Instrument Channel-Orywell High Pressure (PS-64-57A-D)

Tri L vel Settin 1< p<2.5 psig

< 2.5 psig

< 2.5 psig

< 2.5 psig

~Ac

'I h

Remarks 1.

Below trip setting prevents inadvertent operation of containment spray during accident conditions.

l.

Above trip setting in con-junction with low reactor pressure initiates CSS.

Hultiplier relays initiate HPCI.

2.

Hultiplier relay from CSS initiates accident signal.

(15) l.

Above trip setting in conjunction with low reactor pressure initiates LPCI.

l.

Above trip setting, in conjunction with low reactor water level, drywell high

pressure, 120 sec.

delay timer and CSS or RHR pump running, initiates ADS.

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.

2.

In only one trip system.

3.

Not considered in a trip system.

4.

Requires one channel from each physical location (there are 4 locations) in the steam line space.

5.

With diesel power, each RHRS 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 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 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 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 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 3 3.2/4.2-22

NOTES FOR TABLE 2

B ontinued) 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.

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.

Not required to be OPERABLE in the COLD SHUTDOWN CONDITION.

BFN Unit 3 3.2/4.2-23

4 CORE A

D CONTA COOLING SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.B Residual Heat Removal S ste

~RHRS (LPCI and Containment Cooling) 4.5.B Residua eat Removal S ste

~RHRS (LPCI and Containment Cooling)

8. If Specifications 3.5.B.l through 3.5.B.7 are not met, an orderly shutdown shall be initiated and the reactor shall be placed in the COLD SHUTDOWN CONDITION within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

8.

No additional surveillance required.

9.

When the reactor vessel pressure is atmospheric and irradiated fuel is in the reactor vessel, at least one RHR loop with two pumps or two loops with one pump per loop shall be OPERABLE.

The pumps'ssociated diesel generators must also be OPERABLE.

9.

When the reactor vessel pressure is atmospheric, the RHR pumps and valves that are required to be OPERABLE shall be demonstrated to be OPERABLE per Specification 1.0.MM.

10. If the conditions of Specification 3.5.A.5 are met, LPCI and containment cooling are not required.

10.

No additional surveillance required.

ll.

When there is irradiated fuel in the reactor and the reactor is not in the COLD SHUTDOWN CONDITION, 2 RHR pumps and associated heat exchangers and valves on an adjacent unit must be OPERABLE and capable of supplying cross-connect capability except as specified in Specification 3.5.B.12 below.

(Note:

Because cross-connect capability is not a short-term requirement, a

component is not considered inoperable if cross-connect capability can be restored to service within 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.)

11.

The B and D RHR pumps on unit 2 which supply cross-connect capability shall be demonstrated to be OPERABLE per Specification 1.0.MM when the cross-connect capability is required.

BFN Unit 3 3.5/4.5-7

4 CORE A

CO COOLING S S

EMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.B Res dua Heat Remova S stem

~ggSS.

(LPCI and Containment Cooling) 4.5.B Res dual Heat Remova S ste iRRHRd (LPCZ and Containment Cooling) 12.

If one RHR pump or associated heat exchanger located on the unit cross-connection in unit 2 is inoperable for any reason (including valve inoperability, pipe break, etc.),

the reactor may remain in operation for a period not to exceed 30 days provided the remaining RHR pump and associated diesel generator are OPERABLE.

12.

No additional surveillance required.

13.

If RHR cross-connection flow or heat removal capability is lost, the unit may remain in operation for a period not to exceed 10 days unless such capability is restored.

13.

No additional surveillance required.

14.

All recirculation pump discharge valves shall be OPERABLE PRIOR TO STARTUP (or closed if permitted elsewhere in these specifications).

14.

All recirculation pump discharge valves shall be tested for OPERABILITY during any period of COLD SHUTDOWN CONDITION exceeding 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, if OPERABILITY tests have not been performed during the preceding 31 days.

BFN Unit 3 3.5/4.5-8 AMENDMEHTNo. g4 0

4 CORE AND CO COOLI G SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.C RHR Service Water and Emer enc E ui ment Coo in Water S ste s

EECWS Co t ued 4.5.C RHR Service Water and Emer enc ui ment Coo n

Water S stems EECWS Continued 4.

One of the Bl or B2 RHRSW pumps assigned to the RHR heat exchanger supplying the standby coolant supply connection may be inoperable for a period not to exceed 30 days provided the OPERABLE pump is aligned to supply the RHR heat exchanger header and the associated diesel generator and essential control valves are OPERABLE.

4.

No additional surveillance is required.

5.

The standby coolant supply capability may be inoperable for a period not to exceed 10 days.

6. If Specifications 3.5.C.2 through 3.5.C.5 are not
met, an orderly shutdown shall be initiated and the unit placed in the COLD SHUTDOWN CONDITION within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

7.

There shall be at least 2

RHRSW pumps, associated with the selected RHR pumps, aligned for RHR heat exchanger service for each reactor vessel containing irradiated fuel.

BFN Unit 3 3.5/4.5-12 A51ENOMEHT NO. y4 0

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE RE UIREMENTS 3.5.D E ui ment Area Coolers 4.5.D E ui ment Area Coolers 1.

The equipment area cooler associated with each RHR pump and the equipment area cooler associated with each set of core spray pumps (A and C

or B and D) must be OPERABLE at all times when the pump or pumps served by that specific cooler is considered to be OPERABLE.

l.

Each equipment area cooler is operated in conjunction with the equipment served by that particular cooler; therefore, the equipment area coolers are tested at the same frequency as the pumps which they serve.

2.

When an equipment area cooler is not OPERABLE, the pump(s) served by that cooler must be considered inoperable for Technical Specification purposes.

E.

Hi h Pressure Coolant In ection S ste HPCIS E.

Hi h P essure Coo ant In ection S stem HPCIS 1.

The HPCI system shall be OPERABLE whenever there is irradiated fuel in the reactor vessel and the reactor vessel pressure is greater than 150 psig, except in the COLD SHUTDOWN CONDITION or as specified in 3.5.E.2.

OPERABILITY shall be determined within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure reaches 150 psig from a COLD CONDITION, or alternatively PRIOR TO STARTUP by using an auxiliary steam supply.

a.

Simulated Automatic Actuation Test Once/18 months b.

Pump OPERA-BILITY Per Specification 1.0.MM c.

Motor Qper-ated Valve OPERABILITY Per Specification 1.0.MM 1.

HPCI Subsystem testing shall be performed as follows:

d.

Flow Rate at Once/3 normal months reactor vessel operating pressure BFN Unit 3 3.5/4.5-13

.5/4 5

CORE A

CO AI COOLI G SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.E Hi h Pressure Coo ant n ect o

4.5.E Hi h Pressure Coolant In 'ectio 4.5.E.l (Cont'd) e.

Flow Rate at Once/18 150 psig months The HPCI pump shall deliver at least 5000 gpm during each flow rate test.

f.

Verify that Once/Month each valve (manual, power-

operated, or automatic) in the injection flow-path that is not locked, sealed, or otherwise secured in position, is in its correct* position.
2. If the HPCI system is inoperable, the reactor may remain in operation for a period not to exceed 7 days, provided the ADS,. CSS, RHRS (LPCI), and RCICS are OPERABLE.

2.

No additional surveillances are required.

3. If Specifications 3.5.E.l or 3.5.E.2 are not met, an orderly shutdown shall be initiated and the reactor vessel pressure shall be reduced to 150 psig or less within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Except that an automatic valve capable of automatic return to its ECCS position when an ECCS signal is present may be in a position for another mode of operation.

F.

Reactor Core Isolation Coolin F.

Reactor Core Isolation Coolin 1.

The RCICS shall be OPERABLE whenever there is irradiated fuel in the reactor vessel and the reactor vessel pressure is above 150 psig, except in the COLD SHUTDOWN CONDITION or as specified in 3.5.F.2.

OPERABILITY shall 1.

RCIC Subsystem testing shall be performed as follows:

a.

Simulated Auto-Once/18 matic Actuation months Test BFN Unit 3 3.5/4.5-14

4 CORE D

CO t

COO G

S STEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.F eacto Core Iso atio Coolin 4.5.F Reactor Core Isolat on Coolin 3.5.F.l (Cont'd) be determined within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure reaches 150 psig from a COLD CONDITION or alternatively PRIOR TO STARTUP by using an auxiliary steam supply.

4.5.F.l (Cont'd) b.

Pump OPERABILITY

c. Motor-Operated Valve OPERABILITY Per Specifi-cation 1.0.MM Per Specifi-cation 1.0.MM d.

Flow Rate at Once/3 normal reactor months vessel operating pressure e.

Flow Rate at 150 psig Once/18 months The RCIC pump shall deliver at least 600 gpm during each flow test.

2. If the RCICS is inoperable, the reactor may remain in operation for a period not to exceed 7 days if the HPCIS is OPERABLE during such time.
3. If Specifications 3.5.F.l or 3.5.F.2 are not met, an orderly shutdown shall be initiated and the reactor shall be depressurized to less than 150 psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

f.

Verify that Once/Month'ach valve (manual, power-

operated, or automatic) in the injection flowpath that is not locked,
sealed, or other-wise secured in position, is in its correct* position.

2.

No additional surveillances are required.

Except that an automatic valve capable of automatic return to its normal position when a signal is present may be in a position for another mode of operation.

BFN Unit 3 3.5/4.5-15 Al"lENDMEi)TNo

.5 4 CORE AND CO A

COOLING SYS E

S LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.G Automatic De ressur zation 4.5.G Automatic De ressurization Four of the six valves of the Automatic Depressurization System shall be OPERABLE:

1.

During each operating cycle the following tests shall be performed on the ADS:

(1) PRIOR TO STARTUP from a

COLD CONDlTION, or, (2) whenever there is irradiated fuel in the reactor vessel and the reactor vessel pressure is greater than 105 psig, except in the COLD SHUT-DOWN CONDITION or as specified in 3.5.G.2 and 3.5.G.3 below.

a.

A simulated automatic actuation test shall be performed PRIOR TO STARTUP after each refueling outage.

Manual surveillance of the relief valves is covered in 4.6.D.2.

2. If three.of the six ADS valves are known to be incapable of automatic operation, the reactor may remain in operation for a period not to exceed 7 days, provided the HPCI system is OPERABLE. (Note that the pressure relief function of these valves is assured by Section 3.6.D of these specifications and that this specification only applies to the ADS function.) If more than three of the six ADS valves are known to be incapable of automatic operation, an immediate orderly shutdown shall be initiated, with the reactor in a HOT SHUTDOWN CONDITION in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in a COLD SHUTDOWN CONDITION in the following 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br />.

2.

No additional surveillances are required.

3 ~ If Specifications 3.5.G.1 and 3.5.G.2 cannot be met, an orderly shutdown will be initiated and the reactor BFN Unit 3 3.5/4.5-16

4.

CO E

CO COOLI G

SYS E

S LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.G Automat c De ressurization S stem ADS 4.5.G Automatic De ressurization 3.5.G.3 (Cont'd) vessel pressure shall be reduced to 105 psig or less within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

H.

Maintenance of Filled Dischar e

~Pi e

H.

Maintenance of Filled Dischar e

~Pi e

Whenever the core spray systems, LPCI, HPCI, or RCIC are required to be OPERABLE, the discharge piping from the pump discharge of these systems to the last block valve shall be filled.

The following surveillance requirements shall be adhered to assure that the discharge piping of the core spray

systems, LPCI, HPCI, and RCIC are filled:

The suction of the RCIC and HPCI pumps shall be aligned to the condensate storage

tank, and the pressure suppression chamber head tank shall normally be aligned to serve the discharge piping of the RHR and CS pumps.

The condensate head tank may be used to serve the RHR and CS discharge piping if the PSC head tank is unavailable.

The pressure indicators on the discharge of the RHR and CS pumps shall indicate not less than listed below.

1.

Every month and prior to the testing of the RHRS (LPCI and Containment Spray) and core spray systems, the discharge piping of these systems shall be vented from the high point and water flow determined.

2.

Following any period where the LPCI or core spray systems have not been required to be OPERABLE, the discharge piping of the inoperable system shall be vented from the high point prior to the return of the system to service.

Pl-75-20 Pl-75-48 Pl-74-51 P1-74-65 48 psig 48 psig 48 psig 48 psig 3.

Whenever the HPCI or RCIC system is lined up to take suction from the condensate storage

tank, the discharge piping of the HPCI and RCIC shall be vented from the high point of the system and water flow observed on a monthly basis.

4.

When the RHRS and the CSS are required to be OPERABLE, the pressure indicators which monitor the discharge lines shall be monitored daily and the pressure recorded.

BFN Unit 3 3.5/4.5-17

6 4 6

RIMARY SYST OUNDARY LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 4.6.C Coola t Leaka e

l.

a.

Any time irradiated fuel is in the reactor vessel and reactor coolant temperature is above 212'F, reactor coolant leakage into the primary containment from unidentified sources shall not exceed 5 gpm.

In addition, the total reactor coolant system leakage into the primary containment shall not 'exceed 25 gpm.

1.

Reactor coolant system leakage shall be checked by the sump and air sampling system and recorded at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

b.

Anytime the reactor is in RUN mode, reactor coolant leakage into the primary containment from unidentified sources shall not increase by more than 2 gpm averaged over any 24-hour period in which the reactor is in the RUN mode except as defined in 3.6.C.l.c below.

c.

During the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in the RUN mode following

STARTUP, an increase in reactor coolant leakage into the primary containment of >2 gpm is acceptable as long as the requirements of 3.6.C.l.a are met.

BFN Unit 3 3.6/4.6-9 AMEND",ATE/lT NO. Z 0 8

4 RIMARY SYST OUNDARY LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.6.C Coolant Leaka e

4.6.C Coolant Leaka e

2.

Both the sump and air sampling systems shall be OPERABLE during REACTOR POWER OPERATION.

From and after the date that one of these systems is made or found to be inoperable for any

reason, REACTOR POWER OPERATION is permissible only during the succeeding 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for the sump system or 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> for the air sampling system.

2.

With the air sampling system inoperable, grab samples shall be obtained and analyzed at least once every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

The air sampling system may be removed from service for a period of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for calibration, function

testing, and maintenance without providing a temporary monitor.
3. If the condition in 1 or 2 above cannot be met, an orderly shutdown shall be initiated and the reactor shall be placed in the COLD SHUTDOWN CONDITION within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

1.

When more than one relief valve is known to be failed, an orderly shutdown shall be initiated and the reactor depressurized to less than 105 psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

The relief valves are not required to be OPERABLE in the COLD SHUTDOWN CONDITION.

l.

Approximately one-half of all relief valves shall be bench-checked or replaced with a bench-checked valve each operating cycle.

All 13 valves will have been checked or replaced upon the completion of every second cycle.

2.

In accordance with Specification 1.0.MM, each relief valve shall be manually opened until thermocouples and acoustic monitors downstream of the valve indicate steam is flowing from the valve.

BFN Unit 3 3.6/4.6-10

o

3.6/4.6 QASES 3.6.C/4.6.C (Cont'd) suggest a reasonable margin of safety that such leakage magnitude would not result from a crack approaching the critical size for rapid propagation.

Leakage less than the magnitude specified can be detected reasonably in a matter of a few hours utilizing the available leakage detection

schemes, and if the origin cannot be determined in a reasonably short time, the unit should be shut down to allow further investigation and corrective action.

The two gpm limit for coolant leakage rate increase over any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period is a limit specified by the NRC (Reference 2).

This limit applies only duri'ng the RUN mode to avoid being penalized for the expected coolant leakage increase during pressurization.

The total leakage rate consists of all leakage, identified and unidentified, which flows to the drywell floor drain and equipment drain sumps

~

The capacity of the drywell floor sump pump is 50 gpm and the capacity of the drywell equipment sump pump is also 50 gpm.

Removal of 25 gpm from either of these sumps can be accomplished with considerable margin.

References l.

2.

Nuclear System Leakage Rate Limits (BFNP FSAR Subsection 4.10)

Safety Evaluation Report (SER) on IE Bulletin 82-03 D 4 D

Relief Valves To meet the safety basis, 13 relief valves have been installed on the unit with a total capacity of 83.77 percent of nuclear boiler rated steam flow.

The analysis of the worst overpressure transient, (3-second closure of all main steam line isolation valves) neglecting the direct scram (valve position scram) results in a maximum vessel pressure

which, if a neutron flux scram is assumed considering 12 valves
OPERABLE, results in adequate margin to the code allowable overpressure limit of 1,375 psig.

To meet operational

design, the analysis of the plant isolation transient (generator load reject with bypass valve failure to open) shows that 12 of the 13 relief valves limit peak system pressure to a value which is well below the allowed vessel overpressure of 1,375 psig.

Experience in relief and safety valve operation shows that a testing of 50 percent of the valves per year is adequate to detect failures or deteriorations.

The relief and safety valves are benchtested every second operating cycle to ensure that their setpoints are within the g

1 percent tolerance.

The relief valves are tested in place in accordance with Specification 1.0.MM to establish that they will open and pass steam.

BFN Unit 3.6/4.6-30 AMEND%!EAT HO. I4 I

3.6/4.6 BASES 3.6.D/4.6.D (Cont'd)

The requirements established above apply when the nuclear system can be pressurized above ambient conditions.

These requirements are applicable at nuclear system pressures below normal operating pressures because abnormal operational transients could possibly start at these conditions such that eventual overpressure relief would be needed.

However, these transients are much less severe, in terms of pressure, than those starting at rated conditions.

The valves need not be functional when the vessel head is

removed, since the nuclear system cannot be pressurized.

The relief valves are not required to be OPERABLE in the COLD SHUTDOWN CONDITION.

Overpressure protection is provided during hydrostatic tests by two of the relief valves whose relief setting has been established in conformance with ASME Section XI code requirements.

The capacity of one relief valve exceeds the charging capacity of the pressurization source used during hydrostatic testing.

Two relief valves are used to provide redundancy; References 1.

Nuclear System Pressure Relief System (BFNP FSAR Subsection 4.4) 2.

"Protection Against Overpressure" (ASME Boiler and Pressure Vessel

Code,Section III, Article 9) 3.

Browns Ferry Nuclear Plant Design Deficiency ReportTarget Rock Safety-Relief Valves, transmitted by J. E. Gilliland to F. E. Kruesi, August 29, 1973 3.6.E/4.6.E 3~et Pum s Failure of a jet pump nozzle assembly holddown mechanism, nozzle assembly and/or riser, would increase the cross-sectional flow area for blowdown following the design basis double-ended line break.

Also, failure of the diffuser would eliminate the capability to reflood the core to two-thirds height level following a recirculation line break.

Therefore, if a failure

occurred, repairs must be made.

The detection technique is as follows.

With the two recirculation pumps balanced in speed to within g 5 percent, the flow rates in both recirculation loops will be verified by control room monitoring instruments.

If the two flow rate values do not differ by more than 10 percent, riser and nozzle assembly integrity has been verified.

If they do differ by 10 percent or more, the core flow rate measured by the jet pump diffuser differential pressure system must be checked against the core flow rate derived from the measured values of loop flow to core flow correlation. If the difference between measured and derived core flow rate is 10 percent or more (with the derived value higher) diffuser measurements will be taken to define the location within the vessel of failed jet pump nozzle (or riser) and the unit shut down for repairs.

If the potential blowdown flow BFN Unit 3 3.6/4.6-31

3.6/4.6 BASES 3.6.E/4.6.E (Cont'd) area is increased, the system resistance to the recirculation pump is also reduced;

hence, the affected drive pump will "run out" to a substantially higher flow rate (approximately 115 percent to 120 percent for a single nozzle failure). If the two loops are balanced in flow at the same pump speed, the resistance characteristics cannot have 'changed.

Any imbalance between drive loop flow rates would be indicated by the plant process instrumentation.

In

addition, the affected jet pump would provide a leakage path past the core thus reducing the core flow rate.

The reverse flow through the inactive jet pump would still be indicated by a positive differential pressure but the net effect would be a slight decrease (3 percent to 6 percent) in the total core flow measured.

This decrease, together with the loop flow increase, would result in a lack of correlation between measured and derived core flow rate.

Finally, the affected jet pump diffuser differential pressure signal would be reduced because the backflow would be less than the normal forward flow.

A nozzle-riser system failure could also generate the coincident failure of a jet pump diffuser body; however, the converse is not true.

The lack of any substantial stress in the jet pump diffuser body makes failure impossible without an initial nozzle-riser system failure.

3.6.F/4.6.F Recirculation Pum 0 eratio Steady-state operation without forced recirculation will not be permitted for more than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

And the start of a recirculation pump from the natural circulation condition will not be permitted unless the temperature difference between the loop to be started and the core coolant temperature is less than 75 F.

This reduces the positive reactivity insertion to an acceptably low value.

Requiring the discharge valve of the lower speed loop to remain closed until the speed of the faster pump is below 50 percent of its rated speed provides assurance when going from one-to-two pump operation that excessive vibration of the jet pump risers will not occur.

3.6.G/4.6.G Structural Inte it The requirements for the reactor coolant systems inservice inspection program have been identified by evaluating the need for a sampling examination of areas of high stress and highest probability of failure in the system and the need to meet as closely as possible the requirements of Section XI, of the ASME Boiler and Pressure Vessel Code.

The program reflects the built-in limitations of access to the reactor coolant systems.

It is intended that the required examinations and inspection be completed during each 10-year interval.

The periodic examinations are to be done during refueling outages or other extended plant shutdown periods.

BFN Unit 3 3.6/4.6-32

3.6/4.6 BASES 3.6.G/4.6.G (Cont'd)

Only proven nondestructive testing techniques will be used.

More frequent inspections shall be performed on certain circumferential pipe welds as listed in Section 4.6.G.4 to provide additional protection against pipe whip.

These welds were selected in respect to their distance from hangers or supports wherein a failure of the weld would permit the unsupported segments of pipe to strike the drywell wall or nearby auxiliary systems or control systems.

Selection was based on judgment from actual plant observation of hanger and support locations and review of drawings.

Inspection of all these welds during each 10-year inspection interval will result in three additional examinations above the requirements of Section XI of ASME Code.

References l.

Inservice Inspection and Testing (BFNP FSAR Subsection 4.12) 2.

Inservice Inspection of Nuclear Reactor Coolant Systems,Section XI, ASME Boiler and Pressure Vessel Code 3.

ASME Boiler and Pressure Vessel Code,Section III (1968 Edition) 4.

American Society for Nondestructive Testing No. SNT-TC-1A (1968 Edition)

BFN Unit 3 3.6/4.6-33

Enclosure 2

REASON FOR CHANGES, DESCRIPTION, AND JUSTIFICATION BROWNS FERRY NUCLEAR PLANT (BFN)

EASO FOR CHA GES BFN units 1, 2, and 3 technical specifications (TSs) are being changed to:

(1) revise Table 3.2.B and Limiting Conditions for Operation (LCO) 3.5.B.ll, 3.5.E.1, 3.5.F.1, 3.5.G.l, and 3.6.D.l and the bases section for 3.6.D/4.6.D to clarify equipment operability requirements when the reactor is in the cold shutdown condition, (2) revise the maximum operating power level allowed with an inoperable Recirculation Pump Trip (RPT) system(s) from 85 percent to 30 percent

power, and (3) correct two typographical errors in Table 3.2.B.

ESCRIPTIO A

D JUSTIFIC TIO OR E

0 OS C

G S

1.

Clarify equipment operability requirements for when the reactor is in the cold shutdown condition.

a.

Add the following note to Table 3.2.B for the drywell high pressure instruments (PIS-64-57 A-D, PIS-64-58 A-D, and PIS-64-58 E-H for unit 2 and PS-64-57 A-D, PS-64-58 A-D, and PS-64-58 E-H for units 1 and 3):

"18.

Not required to be OPERABLE in the COLD SHUTDOWN CONDITION."

b.

Existing LCO 3.5.B.ll reads in part:

"When there is irradiated fuel in the reactor and the reactor vessel pressure is greater than atmospheric, Revised LCO 3.5.B.ll would read in part:

"When there is irradiated fuel in the reactor and the reactor is not in the COLD SHUTDOWN CONDITION, c.

Existing LCO 3.5.E.1 reads in part:

"The HPCI system shall be OPERABLE

. except as specified in Specification 3.5.E.2 Revised LCO 3.5.E.1 would read in part:

"The HPCI system shall be OPERABLE

. except in the COLD SHUTDOWN CONDITION or as specified in 3.5.E.2 d.

Existing LCO 3.5.F.l reads in part:

"The RCICS shall be OPERABLE except as specified in 3.5.F.2.

Revised LCO 3.5.F.1 would read in part:

"The RCICS shall be operable except in the COLD SHUTDOWN CONDITION or as specified in 3.5.F.2 e.

Existing LCO 3.5.G.1 reads in part:

"Four of the six valves of the Automatic Depressurization System shall be OPERABLE

. except as specified in 3.5.G.2 and 3.5.G.3 below."

Revised LCO 3.5.G.1 would read in part:

"Four of the six valves of the Automatic Depressurization System shall be OPERABLE

. except in the COLD SHUTDOWN CONDITION or as specified in 3.5.G.2 or 3.5.G.3 below."

Page 2 of 5 f.

Existing LCO 3.6.D.1 reads in part:

"When more than one relief valves are known to be failed Revised LCO 3.6.D.l would read in part:

"When more than one relief valve is known to be failed g.

Add the following to LCO 3.6.D.1:

"The relief valves are not required to be OPERABLE in the COLD SHUTDOWN CONDITION."

h.

Add the following paragraph to the bases section for 3.6.D/4.6.D:

"The relief valves are not required to be OPERABLE in the COLD SHUTDOWN CONDITION.

Overpressure protection is provided during hydrostatic tests by two of the relief valves whose relief setting has been established in conformance with ASME section XI code requirements.

The capacity of one relief valve exceeds the charging capacity of the pressurization source used during hydrostatic testing.

Two relief valves are used to provide redundancy."

Table 3.2.B currently requires the drywell high pressure instruments to be operable whenever any core and containment cooling system in TS section 3.5 is operable.

Change (a) above would allow these instruments to be inoperable when the plant is in the cold shutdown condition.

Initiation of these instruments, along with the low reactor pressure instruments, indicates a breach of the nuclear system process barrier within the drywell (steam leak).

With the reactor in the cold shutdown condition (reactor coolant temperature g 212'F and reactor in shutdown or refuel mode),

there is no need to detect steam leaks so it is acceptable for the drywell high pressure instruments to be inoperable.

Additionally, TVA is required by TS section 4.7.A to conduct an Integrated Leakrate Test (ILRT) at certain frequencies.

The ILRT demonstrates the ability of the primary containment to contain the gases released from the primary system during a postulated worst case accident with leakage rates low enough to ensure exposure rates less than the 10 CFR 100 limits.

The test is performed with,the plant in the cold shutdown condition by pressurizing the primary containment (drywell and torus) to design bases accident pressure (49.6 psig) and monitoring pressure and temperature for a prescribed period of time.

From this data, the leakage can be calculated.

The high drywell pressure instruments listed above have a trip level setting of between 1 and 2.5 psig.

Inhibiting these pressure instruments during the ILRT is required to prevent unnecessary Emergency Core Cooling System (ECCS) initiations.

The Residual Heat Removal (RHR) and Core Spray (CS) systems are required to be operable during the test in accordance with LCOs 3.5.A and 3.5.B because reactor pressure is greater than atmospheric.

The reactor low water level instruments (LS-3-58 A-D) are operable and initiate the RHR or CS systems on a low-low reactor water level if necessary.

This ensures that RHR and CS could provide sufficient makeup capacity if required to protect the fuel.

Page 3 of 5 The standby gas treatment and secondary containment systems are also operable during the test and available to contain and filter any radioactive material were it to be released.

TSs currently require that the RHR crosstie be operable with reactor pressure greater than atmospheric (LCO 3.5.B.ll), the High Pressure Coolant Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) systems must be operable with reactor pressure greater the 150 psig, and the Automatic Depressurization System (ADS) and relief valves must be operable with reactor pressure greater than 105 psig.

With the reactor in the cold shutdown condition, vessel pressure is atmospheric except during performance of inservice hydrostatic pressure

tests, inservice leakage tests and ILRT.

Once per inservice inspection interval, the plant is required to perform an inservice hydrostatic pressure test on the reactor vessel and attached piping out to and including the first isolation valve to ensure the retaining capability of the reactor coolant pressure boundary.

The test is performed at pressures (1096 to 1150 psia in the dome) in excess of normal operating pressure (approximately 1020 psia).

An inservice leakage test is requi.red whenever the reactor coolant pressure retaining boundary is breached.

This test is similar to the hydrostatic test, but it is performed at normal operating pressure (approximately 1020 psia).

As described previously, the ILRT is performed by pressurizing the primary containment to 49.6 psig.

As currently written, TSs require the RHR crosstie to be operable for each of these tests and HPCI, RCIC, ADS, and relief valves to be operable for the inservice hydrostatic and leakage tests.

In reality, the ADS is disabled and HPCI and RCIC, both steam turbine driven systems, have no steam supply available during these tests.

These tests are performed in the cold shutdown condition at the end of the refueling outage with fuel loaded and the reactor pressure vessel head installed.

These tests occur when primary system energy is minimal with all control rods inserted.

Because the reactor vessel pressure is greater than atmospheric the RHR and CS systems are required to be operable.

The inservice hydrostatic and leakage tests are performed at or above a

minimum temperature as specified by TS figure 3.6-1.

With the system temperature (approximately 207'F) below the atmospheric pressure boiling point, enthalpy of the bulk fluid is low. If a leak greater than the makeup capacity of the Control Rod Drive (CRD) pump should occur during the test, the system would be depressurized well below the maximum pressure at which the RHR and CS systems could inject to the vessel before water level dropped to an unsafe level.

The available RHR and CS systems are sufficient to preclude fuel uncovering in the event of a leak.

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Page 4 of 5 The RHR crosstie is provided to maintain a long term reactor core and primary containment cooling capability irrespective of primary containment integrity or operability of the RHR system on a given unit.

This is provided in case the torus is breached, flooding the RHR pumps on the affected unit.

However, with the reactor in the cold shutdown condition, there is no high energy potential to breach the torus so the RHR crosstie is not needed.

During inservice hydrostatic testing, ll of the 13 relief valves are disabled by removing the pilot cartridges and blanking the pilot ports.

Overpressure protection is provided by the two remaining relief valves which have their setpoint established in accordance with ASME Section XI.

The relief capacity of one relief valve exceeds the flow capacity of the hydrostatic pressure source.

Two valves are used for redundancy.

These changes are consistent with the General Electric Boiling Water Reactor Standard TSs (NUREG 0123) which requires HPCI (section 3.5.1.c),

RCIC (section 3.7.4),

ADS (section 3.5.1.d),

and relief valves (section 3.4.2.1) to be operable only in the power operation,

startup, and hot shutdown conditions.

2.

Correct the maximum power level allowed with an inoperable RPT system(s).

Existing Table 3.2.B, Note 17 reads:

"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 1 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."

Proposed change to Table 3.2.B, Note 17 would read:

"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."

This change corrects the maximum operating power level allowed with an inoperable RPT system(s) from 85 percent to 30 percent Core Thermal Power (CTP).

Thirty percent CTP is used in the RPT analysis (NED0-24119, "Basis for Installation of Recirculation Pump Trip System for Browns Ferry,"

April 1978, BFN Updated Final Safety Analysis Report (UFSAR) Section 7.9.4.5),

and is conservatively determined to be the maximum power level at which fuel cladding integrity can be assumed during an end of cycle limiting overpressurization event without RPT protection.

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Page 5 of 5 The RPT provides automatic trip of both recirculation pumps after a turbine trip or generator load rejection if reactor power is above approximately 30 percent of rated full load.

The purpose of this trip is to reduce the peak reactor pressure and peak heat flux resulting from transients in which it is postulated that there is a coincident failure of the turbine bypass system.

The recirculation pump trip signal results from either turbine control valve fast closure or turbine stop valve closure.

Reactor scram is also initiated by these signals.

The very rapid reduction in core flow following a recirculation pump trip early in

,these transients reduces the severity of these events because the immediate resultant increase in core voids provides negative reactivity which supplements the negative reactivity from control rod scram.

The proposed change reduces the trip set point from 85 percent to 30 percent and is therefore more conservative than the current operational requirements.

Additionally, the number "1" in Note 17 is being revised to the alphabetic "one" to be consistent with the rest of the note.

3.

Correct two typographical errors in Table 3.2.B.

a.

Correct typographical error in the first entry under "Remarks" in Table 3.2.B (Page 3.2/4.2-14).

Existing entry reads:

"1.

Below trip setting initiated HPCI."

Proposed change to Table 3.2.B would read:

"1.

Below trip setting initiates HPCI."

This change revises the word "initiated" to "initiates" so that this entry will be in the present tense like the other remarks in Table 3.2.B.

b.

Correct typographical error under "Minimum No. Operable Per Trip Sys" column on Table 3.2.B for "RHR (LPCI) Trip System bus power monitor" (page 3.2/4.2-17) Unit 2 only.

The entry in this column should be "l."

This is an omission in the Unit 2 TSs.

The original BFN Unit 2 TSs indicate a "1" in this column as do the current BFN Units 1 and 3 TSs.

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PROPOSED DETERINATION OF NO SIGNIFICANT HAZARDS BROWNS FERRY NUCLEAR PLANT (BFN)

DESCRIPTIO OF PRO OSE TECH ICAL SPECIFICATION S

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T BFN units 1, 2, and 3 technical specifications (TSs) are being changed to:

(1) revise Table 3.2.B and Limiting Conditions for Operation (LCO) 3.5.B.11, 3.5.E.1, 3.5.F.l, 3.5.G.l, and 3.6.D.1 and the bases section for 3.6.D/4.6.D to clarify equipment operability requirements when the reactor is in the cold shutdown condition, (2) revise the maximum operating power level allowed with an inoperable Recirculation Pump Trip (RPT) system(s) from 85 percent to 30 percent

power, and (3) correct two typographical errors in Table 3.2.B.

B SIS FOR PROPOSED 0 SIG I ICA H

RDS CONSIDERATIO DE ERM TIO NRC has provided standards for determining whether a significant hazards consideration exists as stated in 10 CFR 50.92(c).

A proposed amendment, to, an operating license involves no significant hazards considerations if operation of the facility in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated, or (2) create the possibility of a new or different kind of accident from an accident previously evaluated, or (3) involve a significant reduction in a margin of safety.

1.

The proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

Change 1

clarifies equipment operability requirements with the reactor in the cold shutdown condition.

With the reactor in the cold shutdown condition, primary system energy is minimal and the control rods are inserted.

Reactor pressure is normally atmospheric except during performance of inservice hydrostatic tests, inservice leakage tests, and Integrated Leak Rate Tests (ILRT).

This change would inhibit the drywell high pressure instruments which function to detect primary system leaks.

With minimal system energy and no steam generation, this function is not required.

The High Pressure Coolant Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) systems are not required because there is no steam supply to operate them and Residual Heat Removal (RHR) and Core Spray (CS) are operable and capable of providing makeup in case of leaks to protect the fuel from being uncovered.

The Automatic Depressurization System (ADS) is not required for leaks considered possible during the inservice hydrostatic test.

Reactor pressure would decrease fast enough to allow residual heat removal and core spray injection in time to preclude water level decreasing to an unsafe level.

The relief valves are not required to be operable because alternate means of overpressurization protection are provided in the tests.

During inservice hydrostatic testing, ll of the 13 relief valves are disabled by removing the pilot cartridges and blanking the pilot ports.

Overpressure protection is provided by the two remaining relief valves which have their setpoint established in accordance with ASME Section XI.

The RHR crosstie is not required because there is no high energy potential to breach the torus in the cold shutdown condition.

The change is consistent with industry practice and the GE BWR Standard TSs (NUREG 0123).

Page 2 of 2 Change 2 is a more conservative requirement.

The RPT system provides an automatic trip of both recirculation pumps after a turbine trip or a generator load reject.

This reduction in flow increases the core voids and provides immediate negative reactivity to reduce the severity of the transient.

There are two RPT systems.

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 />, reactor power shall be less than 30 percent within four hours (vs. the current 85 percent).

The proposed value of 30 percent power is consistent with the BFN RPT analysis and the BFN Updated Final Safety Analysis Report.

Therefore, this change involves no significant increase in the probability or consequences of an accident previously analyzed.

Change 3 is an administrative change that corrects typographical errors.

2.

The proposed change does not create the possibility of a new or different kind of accident from an accident previously evaluated.

Change l does not involve changes in plant hardware or method of operation from that currently practiced.

The changes are clarifications to TSs to facilitate performance of required TS testing with the reactor in the cold shutdown condition.

The methods of performance are consistent with industry practice.

Change 2 will ensure that when both RPT systems are inoperable or when one RPT system is inoperable more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, reactor power is dropped to a level consistent with the analysis performed for the RPT installation.

Change 3 corrects two typographical errors so the TSs will be more consistent.

3.

The proposed changes do not involve a significant reduction in the margin of safety.

Change l clarifies equipment operability requirements with the reactor in the cold shutdown condition.

Sufficient safety equipment is still available to ensure the fuel remains

covered, even in the event of leaks.

It does not reduce the equipment available to mitigate an accident and as such does not reduce the margin of safety.

Change 2 is more conservative than the current TS.

When the RPT system is inoperable the maximum allowed reactor power will be reduced.

This is consistent with the analysis performed for the RPT installation and the FSAR and does not reduce the margin of safety.

Change 3 is an administrative change which does not reduce the margin of safety.

These changes have been reviewed by TVA.

Based on this review, TVA does not believe the changes present a Significant Hazards Consideration.

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