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*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, SW¹1)Instrument Channel-Reactor Low Mater Level (LIS-3-58A-D, SW¹2)Instrument Channel-Reactor Low Water Level Permissive (LIS-3-184 8 185, SW¹1)Instrument Channel-Reactor Low Water Level (LITS-3-52 and 62, SW¹1)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 ini ti ating si gnal s on ADS.l.Below trip setting prevents inadvertent operation of containment spray during accident condition.
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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, SW¹2)Instrument Channel-Drywell High Pressure (PS-64-58A-D, SW¹1)-Instrument Channel-Drywell High Pressure (PS-64-57A-0)
09035b 901030 PDR ADOCK 05000>59 PNU
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 ini ti ates ace i dent s i gnal.(15)l.Above trip setting in conjunction with low reactor pressure initiates LPCI.l.Above trip setting, in conjunction with 1 ow reac to r 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 hours.If the function is not OPERABLE in 24 hours, take action B.B.Declare the system or component inoperable.
TABLE 3.2.8 INSTRUHENTATION THAT INITATES OR CONTROLS THE CORE AND CONTAINHENT COOLING SYSTEHS Hinimum No.
C.Immediately take action B until power is verified on the trip system.D.No action required;indicators are considered redundant.
Operable Per
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.
~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.
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.
Instrument Channel            >  470" above vessel zero.               1. Hultiplier relays initiate Reactor Low Water Level                                                    RCIC.
Refer to Section 4.5.C for the requirements of a RHRSW pump being inoperable.
Instrument Channel            >  378" above vessel zero.               1. Below  trip setting initiates Reactor Low Water Level                                                    CSS.
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.
(LIS-3-58A-D, SW &#xb9;1)
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.
Hultiplier relays initiate LPCI.
BFN Unit 1 3.2/4.2-24
: 2. Hultiplier relay    from CSS initiates accident signal (15).
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 &#xb9;2)                                                        high pressure, low water level permissive, 120 sec.
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.
Permissive (LIS-3-184 8 185,  SW &#xb9;1)
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 &#xb9;1)                                                  containment spray during accident condition.


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 hours.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 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~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).
4   4 Sr, >~ 4 TABLE  3.2.8 (Continued)
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 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 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 AMEHOMENT g0.y g g
Hinimum No.
/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 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
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.
.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 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.
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.
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
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(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.


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 hours 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 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.
NOTES FOR TABLE      2 B
BFN Unit 1 3.5/4.5-15 AMEHDMEI'tT NO.I 7 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.)
: 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 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~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 hours.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.
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.
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.
Action:
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 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 1 3.6/4.6-9 AMENDMENT NO.I 3 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.
A. Repair in 24 hours.        If the    function is not    OPERABLE  in  24 hours, take action B.
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.
B. Declare the system or component inoperable.
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.
C. Immediately take action        B  until    power  is verified  on the  trip system.
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.
D. No  action required; indicators are considered redundant.
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.
: 2. In only  one  trip  system.
The relief valves are benchtested every second operating cycle to ensure that their setpoints are within the g 1 percent tolerance.
: 3. Not considered    in a  trip  system.
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.
: 4. Requires one channel from each physical location (there are 4 locations) in the steam line space.
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.
: 5. With diesel power, each        RHRS pump    is  scheduled to start immediately and each  CSS  pump  is sequenced    to start about      7 sec. later.
The valves need not be functional when the vessel head is removed, since the nuclear system cannot be pressurized.
: 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.,
The relief valves are not required to be OPERABLE in the COLD SHUTDOWN CONDITION.
at which time the      full complement      of  CSS  and RHRS pumps  would be operating.
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.
: 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.
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.
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.
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 Report-Target 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
: 8. Note 1 does not apply to        this item.
~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.
: 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.
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.
BFN                                             3.2/4.2-23 Unit 1
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 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.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 ti on 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~Ainn A'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.Hul tiplier relay from CSS ini ti ates ace i dent s i gnal.(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 I r i~!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.Moni tors 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 hours.If the function is not OPERABLE in 24 hours, 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 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 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 hours.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 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 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 hours, if OPERABILITY tests have not been performed during the preceding 31 days.BFN Unit 2 3.5/4.5-8 AMENDMENT NO.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 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 AMfNDMEHT NO.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 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 4.5 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 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 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 hours 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 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 AMENOMBlT WO.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 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 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.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 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 AMENDMEHT tl0.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.
NOTES FOR   ABLE    2 B    ont'd)
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.
: 10. Only one  trip  system  for  each cooler fan.
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.
: 11. In only  two of the four 4160-V shutdown boards.       See  note 13.
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.
: 12. In only  one  of the four    4160-V shutdown boards. See  note 13.
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.
: 13. An emergency 4160-V shutdown board        is considered  a  trip system.
The relief valves are benchtested every second operating cycle to ensure that their setpoints are within the+1 percent tolerance.
: 14. RHRSW pump  would be inoperable. Refer to Section 4.5.C        for the requirements of a RHRSW pump being inoperable.
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 AMENOMENT 50.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.
: 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.
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.
: 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.
The valves need not be functional when the vessel head is removed, since the nuclear system cannot be pressurized.
: 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.
The relief valves are not required to be OPERABLE in the COLD SHUTDOWN CONDITION.
: 18. 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.
BFN                                        3.2/4.2-24 Unit 1
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 Report-Target 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.
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
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.
        ~RfLRRS  . (LPCI and Containment                                  ~RHRS    iLPCP and Containment Cooling)                                                          Cooling)
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.
: 8. If Specifications    3.5.B.1                              8. No  additional surveillance through 3.5.B.7 are not met,                                    required.
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.
an orderly shutdown shall be initiated and the reactor shall  be placed    in the COLD SHUTDOWN CONDITION within  24 hours.
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*
: 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.
3.2/4.2-23 3.5/4.5-7 3.5/4.5-8*
diesel generators pumps'ssociated must also be OPERABLE.
3.5/4.5-12*
: 10. If the conditions of                                      10. No  additional surveillance Specification 3.5.A.5 are met,                                   required.
3.5/4.5-13 3.5/4.5-14 3.5/4.5-15*
LPCI and containment    cooling are not required.
3.5/4.5-16 3.5/4.5-17*
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.
3.6/4.6-9*
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.)
3.6/4.6-10 3.6/4.6-30*
BFN                                           3.5/4.5-7 Unit  1
3.6/4.6-31 3.6/4.6-32*
 
3.6/4.6-33*
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
        ~RHRS    (LPCI and Containment                    ~RHRS    (LPCI and Containment Cooling)                                          Cooling)
: 12. If one  RHR pump  or associated            12. No  additional surveillance required.
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.
: 13. If  RHR cross-connection    flow or          13. No  additional surveillance heat removal capability is lost,                  required.
the unit may remain in operation for a period not to exceed 10 days unless such capability is restored.
: 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.
3.5/4.5-8 AMENDMEN [f0. g G 9 BFN Unit  1
 
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. One of the Dl  or D2 RHRSW                4. No additional surveillance pumps assigned  to the RHR                      is required.
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.
: 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                                          3.5/4.5-12                AMEHOMENT g0. ygg Unit  1
 
    /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                  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.
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.
E. Hi h Pressure    Coolant In ection              E. H  h Pressure    Coolant S  stem    HPC S                                      In ection    S stem  HPCIS
: 1. The HPCI system    shall be                    1. HPCI Subsystem  testing OPERABLE  whenever there is                          shall be performed as irradiated fuel in the                                follows:
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
: d. Flow Rate at    Once/3 normal          months reactor vessel operating pressure BFN                                            3.5/4.5-13 Unit  1
 
~ ~
.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
: 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,                    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,
: 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.
shall be depressurized to less than 150 psig within                      2. No  additional surveillances 24 hours.                                          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'tT NO. 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                 1. During each operating the Automatic                                  cycle the following Depressurization System                        tests shall be performed shall  be OPERABLE:                            on the ADS:
(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 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.
DOWN  CONDITION  or as specified in 3.5.G.2  and 3.5.G.3 below.
: 2. If  three of the six ADS                    2. No additional surveillances valves are known to be                          are required.
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.
: 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
 
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.
H. Maintenance   o  F    ed Dischar e        H. Ma  ntenance    of Filled Dischar  e
          ~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:
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.
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.
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.
: 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                                        3.5/4.5-17 Unit  1
 
4   PRIMARY SYSTE    0    ARY LIMITING CONDITIONS  FOR OPERATION                SURVEILLANCE RE UIREMENTS 4.6.C. Coo ant  eaka e
: 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.
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.
: 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                                      3.6/4.6-9              AMENDMENT NO. I3 7 Unit  1
 
4    PRIMARY SYSTE    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                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.
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.
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          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.
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.
: 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                                      3.6/4.6-10 Unit  1
 
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
: 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                                        3.6/4.6-30          AMENDMENT NO g~O Unit  1
 
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 Report      Target 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                                          3.6/4.6-31 Unit  1
 
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 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.
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.
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
 
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                                      3.6/4.6-33 Unit 1
 
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*
3.2/4.2-17                      3.2/4.2-17 3.2/4.2-23                      3.2/4.2-23+
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*
3.5/4.5-12                      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*
3.5/4.5-16                      3.5/4.5-16 3.5/4.5-17                      3.5/4.5-17*
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*
3.6/4.6-31                      3.6/4.6-31 3.6/4.6-32.                    3.6/4.6-32*
3.6/4.6-33                      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.
~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.
(LIS-3-58A-D)
Instrument Channel          > 470" above vessel  zero. A              1. Multiplier relays initiate Reactor Low Water Level                                                          RCIC.
(LIS-3-58A-D) 2              Instrument Channel          > 378" above vessel  zero. A              1. Below  trip setting initiates Reactor Low Water Level                                                          CSS.
(LS-3-58A-0)
Hultiplier relays initiate LPCI.
: 2. Multiplier relay from CSS initiates accident signal (15).
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.
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          > 544" above vessel  zero. A              1. Below  trip setting  permissive Reactor Low Water Level                                                          for initiating signals    on ADS.
Permissive (LIS-3-184, 185)
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.
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
                                                                  ~Ainn T~ri S  i 2(18)
Func i n Instrument Channel-Drywell High Pressure 1<
Tri  L v  1 p<2.5 psig in A'emarks
: l. Below  trip setting prevents inadvertent operation of (PIS-64-58 E-H)                                            containment spray during accident conditions.
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.
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(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.
 
                                                                                                            ~ ~
TABLE    3.2.B (Continued)
Hinimum No.
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.
(PIS-68-95, 96)
Instrument Channel-        230  psig +    15              1. Recirculation discharge valve Reactor Low Pressure                                            actuation.
(PS-3-74 A 5 B)
(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.
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.
RHRSW  Al,  B3, Cl, and 03  27<  t  <  29 sec.              1. With normal power Timers                                                      2. One per pump
 
TABLE  3.2.B (Continued)
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.
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.
delay timer, and RHR or CSS pumps running, initiates  AOS.
Instrument Channel-          100 +10  psig                      l. Below  trip setting  defers ADS RHR  Discharge Pressure                                            actuation.
Instrument Channel            185 +10  psig                      l. Below  trip setting  defers AOS CSS Pump  Discharge Pressure                                      actuation.
1(3)  Core Spray Sparger to        2  psid +0.4                          Alarm to detect core sparger Reactor Pressure Vessel d/p                                        pipe break.
RHR  (LPCI) Trip System bus  N/A                                1. Monitors  availability of power monitor Core Spray Trip System bus power monitor N/A                                1.
power to Moni tors power logic systems.
avail abil i ty of to logic systems.
I
                                                                                                                    ~
ADS  Trip  System bus power  N/A                                1. Monitors  availability of monitor                                                            power  to logic systems and  valves.
 
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 hours.      If the  function is not  OPERABLE  in  24 hours, 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                                              3.2/4.2-23 Unit 2
 
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 reduction shall be initiated for  more than 72 hours, an and reactor power shall be orderly less than power 30  percent within four hours.
: 18. Not required to be    OPERABLE  in the COLD SHUTDOWN CONDITION.
BFN                                      3.2/4.2-24 Unit  2
: 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
        ~RHRS    (LPCI and Containment                                    ~RHRS    (LPCI and Containment Cooling)                                                          Cooling)
: 8. If Specifications    3.5.B.l                                8. No  additional surveillance through 3.5.B.7 are not met,                                      required.
an  orderly shutdown shall                    be initiated  and the reactor shall be placed in the COLD SHUTDOWN CONDITION within  24  hours.
: 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.
diesel generators pumps'ssociated must also be OPERABLE.
: 10. If the conditions of                                        10. No  additional surveillance Specification 3.5.A.5 are met,                                    required.
LPCI and containment      cooling are not required.
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.
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.)
BFN                                        3.5/4.5-7 Unit  2
 
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
        ~RHRS  (LPCI and Containment                        PHRS (LPCI    and Containment Cooling)                                            Cooling)
: 12. If three  RHR pumps  or associated          12. No  additional surveillance required.
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            13. No  additional surveillance heat removal capability is lost,                    required.
the unit may remain in operation for a period not to exceed 10 days unless such capability is restored.
: 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.
AMENDMENT NO. X6  9 BFN                                            3.5/4.5-8 Unit  2
 
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. Three  of the Dl,  D2,  Bl,  B2            4. No additional surveillance RHRSW pumps assigned    to  the                  is required.
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.
: 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                                              3.5/4.5-12            AMfNDMEHTNO. I6 9 Unit  2
 
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                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.
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.
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:
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.
: d. Flow Rate at    Once/3 normal          months reactor vessel operating pressure BFN                                    3.5/4.5-13 Unit  2
 
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.
: 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                    are required.
may 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 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.
hours.
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  2
 
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
: 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,                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,
: 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.
shall be depressurized to less than 150 psig within                2. No  additional surveillances 24  hours.                                    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.
AMENOMBlT WO. X V 6 BFN                                          3.5/4. 5-15 Unit  2
 
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                1. During each operating the Automatic                                    cycle the following Depressurization System                          tests shall be performed shall  be OPERABLE:                              on the ADS:
(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 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.
DOWN  CONDITION  or as specified in 3.5.G.2 and 3.5.G.3 below.
: 2. If  three of the six ADS                      2. No  additional surveillances valves are known to be                            are required.
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.
: 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
 
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.
H. Maintenance      of Filled Dischar    e        H. aintenance of    F  lied Dischar e
        ~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:
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.
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.
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.
: 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                                          3.5/4.5-17 Unit  2
 
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
                ~
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.
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.
: 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                                          3.6/4.6-9 AMENDMEHT tl0. I~ 3 Unit  2
 
4    PRIMARY SYSTE      OUNDARY LIMITING CONDITIONS  FOR OPERATION              SURVEILLANCE REQUIREMENTS 4.6.C  Coo ant Leaka  e
: 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.
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.
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        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.
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.
: 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                                      3.6/4.6-10 Unit  2
 
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                                        3.6/4.6-30 AMENOMENT 50. IP g Unit  2
 
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 Report      Target 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                                      3.6/4.6-31 Unit  2
 
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 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).
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.
BFN                                      3.6/4.6-32 Unit  2
 
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                                      3.6/4.6-33 Unit 2
 
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*
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*
3.5/4.5-12                    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*
3.5/4.5-16                    3.5/4.5-16 3.5/4.5-17                    3.5/4.5-17*
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*
3.6/4.6-31                    3.6/4.6-31 3.6/4.6-32                    3.6/4.6-32*
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.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, 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.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 hours.If the function is not OPERABLE in 24 hours, 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
TABLE 3.2.8 INSTRUMENTATION THAT INITIATES OR CONTROLS THE CORE AND CONTAINHENT COOLING SYSTEHS Hinimum No.
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 .
Instrument Channel          > 470" above vessel  zero.                    1. Multiplier relays initiate Reactor Low Mater Level                                                          RCIC.
Instrument Channel-          > 378" above vessel  zero. A              1. Below  trip setting initiates Reactor Low Water Level                                                          CSS.
(LIS-3-58A-D, SW01)
Hultiplier relays initiate LPCI.
: 2. Hultiplier relay from CSS initiates accident signal (15).
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.
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


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.
PROPOSED  DETERINATION OF NO SIGNIFICANT HAZARDS BROWNS FERRY NUCLEAR PLANT (BFN)
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.
DESCRIPTIO    OF PRO OSE    TECH ICAL SPECIFICATION    S A  D    T BFN  units 1, 2, and 3 technical specifications (TSs) are being changed to:
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.
(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.
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.
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.
Additionally, TVA is required by TS section 4.7.A to conduct an Integrated Leakrate Test (ILRT)at certain frequencies.
: 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.
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.
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).
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.
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.
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.
Therefore, this change involves no significant increase in the probability or consequences of an accident previously analyzed.
I'S 4 I I V nl 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.
Change 3  is an administrative    change that corrects typographical errors.
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. 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.Correct the maximum power level allowed with an inoperable RPT system(s).
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.
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.
Change 3   corrects  two typographical errors so the    TSs  will be  more consistent.
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: "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.
: 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.
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.
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.
l II s 4 o ,I 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.
Change 3  is an administrative    change which does not reduce the margin        of safety.
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).
These changes have been reviewed by TVA. Based on this review, TVA does not believe the changes present a Significant Hazards Consideration.
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.
0 r I 1 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: (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.
hl J
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.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.
      ~a
The changes are clarifications to TSs to facilitate performance of required TS testing with the reactor in the cold shutdown condition.
          ~ I}}
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 hours, 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.
hl J~/~a~I}}

Revision as of 23:35, 21 October 2019

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
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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 3.2/4.2-14 3.2/4.2-15 3.2/4.2-15 3.2/4.2-23 3.2/4.2-23*

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*

3.5/4.5-12 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*

3.5/4.5-16 3.5/4.5-16 3.5/4.5-17 3.5/4.5-17*

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*

3.6/4.6-31 3.6/4.6-31 3.6/4.6-32 3.6/4.6-32*

3.6/4.6-33 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 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.

Instrument Channel > 470" above vessel zero. 1. Hultiplier relays initiate Reactor Low Water Level RCIC.

Instrument Channel > 378" above vessel zero. 1. Below trip setting initiates Reactor Low Water Level CSS.

(LIS-3-58A-D, SW ¹1)

Hultiplier relays initiate LPCI.

2. Hultiplier relay from CSS initiates accident signal (15).

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.

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.

Permissive (LIS-3-184 8 185, SW ¹1)

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.

4 4 Sr, >~ 4 TABLE 3.2.8 (Continued)

Hinimum No.

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.

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 ¹2) pressure initiates CSS.

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 ¹1)- reactor pressure initiates LPCI.

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.

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 3.2/4.2-23 Unit 1

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 3.2/4.2-24 Unit 1

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

~RfLRRS . (LPCI and Containment ~RHRS iLPCP and Containment Cooling) Cooling)

8. If Specifications 3.5.B.1 8. No additional surveillance through 3.5.B.7 are not met, required.

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 />.

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.

diesel generators pumps'ssociated must also be OPERABLE.

10. If the conditions of 10. No additional surveillance Specification 3.5.A.5 are met, required.

LPCI and containment cooling are not required.

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.

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.)

BFN 3.5/4.5-7 Unit 1

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

~RHRS (LPCI and Containment ~RHRS (LPCI and Containment Cooling) Cooling)

12. If one RHR pump or associated 12. No additional surveillance required.

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.

13. If RHR cross-connection flow or 13. No additional surveillance heat removal capability is lost, required.

the unit may remain in operation for a period not to exceed 10 days unless such capability is restored.

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 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.

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

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. One of the Dl or D2 RHRSW 4. No additional surveillance pumps assigned to the RHR is required.

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.

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 3.5/4.5-12 AMEHOMENT g0. ygg Unit 1

/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 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.

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.

E. Hi h Pressure Coolant In ection E. H h Pressure Coolant S stem HPC S In ection S stem HPCIS

1. The HPCI system shall be 1. HPCI Subsystem testing OPERABLE whenever there is shall be performed as irradiated fuel in the follows:

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 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 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

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

~ ~

.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 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 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

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, 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,
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.

shall be depressurized to less than 150 psig within 2. No additional surveillances 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. 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'tT NO. 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 1. During each operating the Automatic cycle the following Depressurization System tests shall be performed shall be OPERABLE: on the ADS:

(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 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.

DOWN CONDITION or as specified in 3.5.G.2 and 3.5.G.3 below.

2. If three of the six ADS 2. No additional surveillances valves are known to be are required.

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 />.

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

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 H. Ma ntenance of Filled Dischar e

~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:

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.

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.

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.

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 3.5/4.5-17 Unit 1

4 PRIMARY SYSTE 0 ARY LIMITING CONDITIONS FOR OPERATION SURVEILLANCE RE UIREMENTS 4.6.C. Coo ant eaka e

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 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

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.

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 3.6/4.6-9 AMENDMENT NO. I3 7 Unit 1

4 PRIMARY SYSTE 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 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.

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.

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 hours.

D. Relief Valves D. Relief Valves

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 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The each operating cycle.

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.

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 3.6/4.6-10 Unit 1

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

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 Report Target 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 3.6/4.6-31 Unit 1

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 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.

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 3.6/4.6-32 Unit

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 3.6/4.6-33 Unit 1

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*

3.2/4.2-17 3.2/4.2-17 3.2/4.2-23 3.2/4.2-23+

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*

3.5/4.5-12 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*

3.5/4.5-16 3.5/4.5-16 3.5/4.5-17 3.5/4.5-17*

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*

3.6/4.6-31 3.6/4.6-31 3.6/4.6-32. 3.6/4.6-32*

3.6/4.6-33 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.

~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.

(LIS-3-58A-D)

Instrument Channel > 470" above vessel zero. A 1. Multiplier relays initiate Reactor Low Water Level RCIC.

(LIS-3-58A-D) 2 Instrument Channel > 378" above vessel zero. A 1. Below trip setting initiates Reactor Low Water Level CSS.

(LS-3-58A-0)

Hultiplier relays initiate LPCI.

2. Multiplier relay from CSS initiates accident signal (15).

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.

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 > 544" above vessel zero. A 1. Below trip setting permissive Reactor Low Water Level for initiating signals on ADS.

Permissive (LIS-3-184, 185)

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.

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

~Ainn T~ri S i 2(18)

Func i n Instrument Channel-Drywell High Pressure 1<

Tri L v 1 p<2.5 psig in A'emarks

l. Below trip setting prevents inadvertent operation of (PIS-64-58 E-H) containment spray during accident conditions.

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.

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(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.

~ ~

TABLE 3.2.B (Continued)

Hinimum No.

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.

(PIS-68-95, 96)

Instrument Channel- 230 psig + 15 1. Recirculation discharge valve Reactor Low Pressure actuation.

(PS-3-74 A 5 B)

(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.

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.

RHRSW Al, B3, Cl, and 03 27< t < 29 sec. 1. With normal power Timers 2. One per pump

TABLE 3.2.B (Continued)

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.

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.

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

Instrument Channel- 100 +10 psig l. Below trip setting defers ADS RHR Discharge Pressure actuation.

Instrument Channel 185 +10 psig l. Below trip setting defers AOS CSS Pump Discharge Pressure actuation.

1(3) Core Spray Sparger to 2 psid +0.4 Alarm to detect core sparger Reactor Pressure Vessel d/p pipe break.

RHR (LPCI) Trip System bus N/A 1. Monitors availability of power monitor Core Spray Trip System bus power monitor N/A 1.

power to Moni tors power logic systems.

avail abil i ty of to logic systems.

I

~

ADS Trip System bus power N/A 1. Monitors availability of monitor power to logic systems and valves.

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 3.2/4.2-23 Unit 2

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 reduction shall be initiated for more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, an and reactor power shall be orderly less than power 30 percent within four hours.
18. Not required to be OPERABLE in the COLD SHUTDOWN CONDITION.

BFN 3.2/4.2-24 Unit 2

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

~RHRS (LPCI and Containment ~RHRS (LPCI and Containment Cooling) Cooling)

8. If Specifications 3.5.B.l 8. No additional surveillance through 3.5.B.7 are not met, required.

an orderly shutdown shall be initiated and the reactor shall be placed in the COLD SHUTDOWN CONDITION within 24 hours.

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.

diesel generators pumps'ssociated must also be OPERABLE.

10. If the conditions of 10. No additional surveillance Specification 3.5.A.5 are met, required.

LPCI and containment cooling are not required.

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.

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 />.)

BFN 3.5/4.5-7 Unit 2

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

~RHRS (LPCI and Containment PHRS (LPCI and Containment Cooling) Cooling)

12. If three RHR pumps or associated 12. No additional surveillance required.

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 13. No additional surveillance heat removal capability is lost, required.

the unit may remain in operation for a period not to exceed 10 days unless such capability is restored.

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 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.

AMENDMENT NO. X6 9 BFN 3.5/4.5-8 Unit 2

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. Three of the Dl, D2, Bl, B2 4. No additional surveillance RHRSW pumps assigned to the is required.

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.

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 3.5/4.5-12 AMfNDMEHTNO. I6 9 Unit 2

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 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.

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.

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:

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 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 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.

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

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.
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 are required.

may 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 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.

hours.

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 2

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 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 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

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, 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,
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.

shall be depressurized to less than 150 psig within 2. No additional surveillances 24 hours. 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.

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

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 1. During each operating the Automatic cycle the following Depressurization System tests shall be performed shall be OPERABLE: on the ADS:

(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 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.

DOWN CONDITION or as specified in 3.5.G.2 and 3.5.G.3 below.

2. If three of the six ADS 2. No additional surveillances valves are known to be are required.

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 />.

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

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 H. aintenance of F lied Dischar e

~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:

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.

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.

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.

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 3.5/4.5-17 Unit 2

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

~

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 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

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.

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 3.6/4.6-9 AMENDMEHT tl0. I~ 3 Unit 2

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

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 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

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.

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 hours.

D. Re ef Valves

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 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The each operating cycle.

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.

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 3.6/4.6-10 Unit 2

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 3.6/4.6-30 AMENOMENT 50. IP g Unit 2

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 Report Target 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 3.6/4.6-31 Unit 2

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 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).

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 3.6/4.6-32 Unit 2

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 3.6/4.6-33 Unit 2

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*

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*

3.5/4.5-12 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*

3.5/4.5-16 3.5/4.5-16 3.5/4.5-17 3.5/4.5-17*

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*

3.6/4.6-31 3.6/4.6-31 3.6/4.6-32 3.6/4.6-32*

3.6/4.6-33 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 Tri Level ettin Action Remarks Instrument Channel- > 470" above vessel zero. 1. Below trip setting initiates Reactor Low Water Level HPCI .

Instrument Channel > 470" above vessel zero. 1. Multiplier relays initiate Reactor Low Mater Level RCIC.

Instrument Channel- > 378" above vessel zero. A 1. Below trip setting initiates Reactor Low Water Level CSS.

(LIS-3-58A-D, SW01)

Hultiplier relays initiate LPCI.

2. Hultiplier relay from CSS initiates accident signal (15).

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.

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.

Permissive (LIS-3-184 4 185, SM41)

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.

TABLE 3.2.B (Continued)

Hinimum No.

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.

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, SW¹2) pressure initiates CSS.

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, SW¹1) reactor pressure initiates LPCI.

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.

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 hours, 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 3.2/4.2-22 Unit 3

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 3.2/4.2-23 Unit 3

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

~RHRS (LPCI and Containment ~RHRS (LPCI and Containment Cooling) Cooling)

8. If Specifications 3.5.B.l 8. No additional surveillance through 3.5.B.7 are not met, required.

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 />.

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.

must also be OPERABLE.

10. If the conditions of 10. No additional surveillance Specification 3.5.A.5 are met, required.

LPCI and containment cooling are not required.

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.

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 />.)

BFN 3.5/4.5-7 Unit 3

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

~ggSS. (LPCI and Containment iRRHRd (LPCZ and Containment Cooling) Cooling)

12. If one RHR pump or associated 12. No additional surveillance heat exchanger located required.

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.

13. If RHR cross-connection flow or 13. No additional surveillance heat removal capability is lost, required.

the unit may remain in operation for a period not to exceed 10 days unless such capability is restored.

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 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, OPERABILITY tests have if not been performed during the preceding 31 days.

BFN 3.5/4.5-8 AMENDMEHT No. g4 0 Unit 3

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. One of the Bl or B2 RHRSW 4. No additional surveillance pumps assigned to the RHR is required.

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.

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

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 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.

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.

E. Hi h Pressure Coolant In ection E. Hi h P essure Coo ant S ste HPCIS In ection S stem HPCIS

1. The HPCI system shall be 1. HPCI Subsystem testing OPERABLE whenever there is shall be performed as irradiated fuel in the follows:

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

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

.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 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 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.

hours.

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 3

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

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, 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,
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.

shall be depressurized to less than 150 psig within 2. No additional surveillances 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. 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 3.5/4.5-15 Al"lENDMEi)TNo Unit 3

.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:

(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.

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.

DOWN CONDITION or as specified in 3.5.G.2 and 3.5.G.3 below.

2. If three .of the six ADS 2. No additional surveillances valves are known to be are required.

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 />.

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

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 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 H. Maintenance of Filled Dischar e

~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:

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.

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.

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.

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 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

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 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

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.

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 3.6/4.6-9 AMEND",ATE/ lT NO. Z0 8 Unit 3

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 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 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

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.

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 hours.
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 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The each operating cycle.

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.

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 3.6/4.6-10 Unit 3

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. Nuclear System Leakage Rate Limits (BFNP FSAR Subsection 4.10)
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.

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

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 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.

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 3.6/4.6-31 Unit 3

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 3.6/4.6-32 Unit 3

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 3.6/4.6-33 Unit 3

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 A D 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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