ML18033B436
| ML18033B436 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 07/13/1990 |
| From: | TENNESSEE VALLEY AUTHORITY |
| To: | |
| Shared Package | |
| ML18033B435 | List: |
| References | |
| TVA-BFN-TS-290, NUDOCS 9007180126 | |
| Download: ML18033B436 (45) | |
Text
ENCLOSURE 1
PROPOSED TECHNICAL SPECIFICATION BROMNS FERRX NUCLEAR PLANT UNIT 2 (TVA BFN TS 290) 9007180126 900713 PDR ADOCK 05000260 P
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UNIT 2 EFFECTIVE PAGE LIST REMOVE INSERT 3.2/4.2-18 3.2/4.2-19 3.2/4.2-22 3.2/4.2-23 3.2/4.2-24 3.2/4.2-44 3.2/4.2-45 3.2/4.2-46 3.2/4.2-47 3.2/4.2-67 3.2/4.2-68 3.2/4.2-69 3.2/4.2-70 3.2/4.2-71 3.2/4.2-72 3.2/4.2-73 3.2/4.2-18
.3.2/4.2-19 3.2/4.2-22*
3.2/4.2-22a 3.2/4.2-23 3.2/4.2-24*
3.2/4.2-44*
3.2/4.2-45 3.2/4.2-46 3.2/4.2-47 3.2/4.2-67 3.2/4.2-68*
3.2/4.2-69*
3.2/4.2-70*
3.2/4.2-71*
3.2/4.2-72*
3.2/4.2-73*
3.2/4.2-73a*
- Denotes overleaf or spillover page.
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ENCLOSURE 2
SUMMARY
OF CHANGES (UNIT 2) 1.
Revise Table 3.2.B.
a.
Delete the following from Table 3.2.B.
"Instrument Channel RCIC Steam Line Space High Temperature" on page 3.2/4.2-18 "Instrument Channel HPCI Steam Line Space High Temperature" on page 3.2/4.2-19 b.
Add the following to Table 3.2.B.
"Minimum No.
Operable Per Tri S
s 1)
Function Allowable Value Action Remarks RCIC Steam Line Space Torus Area High Temperature 155 F.
l.
Above trip setting isolates RCIC system and trips RCIC turbine.
RCIC Steam Line Space
< 180'.
RCIC Pump Room Area High Temperature 1.
Above trip setting isolates RCIC system and trips RCIC turbine.
HPCI Steam Line Space Torus Area High Temperature 180'.
l.
Above trip setting isolates HPCI system and trips HPCI turbine.
HPCI Steam Line Space
< 200'.
HPCI Pump Room Area High Temperature 1.
Above trip setting isolates HPCI system and trips HPCI turbine."
c.
Add a new note 1.E to the notes for Table 3.2.B "E.
Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> restore the inoperable channel(s) to OPERABLE status or place the inoperable channel(s) in the tripped condition."
d.
Delete note 4 of the notes for Table 3.2.B.
Existing note 4 reads:
"4.
Requires one channel from each physical location (there are 4
locations) in the steam line space."
Revised note 4 would read:
Deleted"
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Enclosure 2
Page 2 of 3 2.
Revise Table 4.2.B a.
Delete the following from Table 4.2.B on page 3.2/4.2-46.
"Instrument Channel RCIC Steam Line Space High Temperature" b.
Delete the following from Table 4.2.B on page 3.2/4.2-47.
"Instrument Channel HPCI Steam Line Space High Temperature" c.
Add the following to Table 4.2.B "Function Functional Test Calibration Instrument Check RCIC Steam Line Space Torus Area High Temperature Once/3 months none RCIC Steam Line Space RCIC Pump Room Area High Temperature'nce/3 months none HPCI Steam Line Space Torus Area High Temperature Once/3 months none HPCI Steam Line Space HPCI Pump Room Area High Temperature Once/3 months none" 3.
Revise bases section 3.2.
F Existing bases reads in part on page 3.2/4.2-67:
.The HPCI trip settings of 90 psi for high flow and 200' for high temperature are such that core uncovery is prevented and fission product release is within limits.
The RCIC high flow and temperature instrumentation are arranged the same as that for the HPCI.
The trip setting of 450" HzO for high flow and 200' for temperature are based on the same criteria as the HPCI.
Revised bases section 3.2 would read in part:
Each trip system consists of two channels.
Each channel contains one temperature switch located in the pump room and three temperature switches located in the torus area.
The RCIC high flow and high area temperature sensing instrument channels are arranged
'in the same manner as the HPCI system.
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- Enclosure 2
Page 3 of 3 The HPCI high steam flow trip setting of 90 psid and the RCIC high steam flow trip setting of 450" H~O have been selected such that the trip setting is high enough to prevent spurious tripping during pump startup but low enough to prevent core uncovery and maintain fission product releases within 10 CFR 100 limits.
The HPCI and RCIC steam line space temperature switch trip settings are high enough to prevent spurious isolation due to normal temperature excursions in the vicinity of the steam supply piping.
Additionally, these trip settings ensure that the primary containment isolation steam supply valves isolate a break within an acceptable time period to prevent core uncovery and maintain fission product releases within 10 CFR 100 limits.
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ENCLOSURE 3
REASONS AND JUSTIFICATION FOR THE CHANGES Reasons for the Chan es HPCI and RCIC steam line space high temperature isolations specified in Tables 3.2.B and 4.2.B are provided to ensure that automatic closure of each system's primary containment isolation valves occurs if a system's steam line breaks to prevent the excessive loss of reactor coolant and the release of significant amounts of radioactive material from the nuclear system process barrier.
TVA has utilized computer modeling techniques to predict the temperature response of various reactor building zones to high energy line breaks (HELBs).
The results indicate that temperatures below the present 200' technical specification value could be present for various HPCI and RCIC line break scenarios.
Therefore, the current 200' value in the technical specifications must be revised.
A summary of the proposed changes to Tables 3.2.B and 4.2.B and to bases section 3.2 is provided by Enclosure 2.
Justification for the Chan es The HPCI system is provided to limit fuel cladding temperature in the event of a small break in the nuclear system and loss of coolant which does not result in rapid depressurization of the reactor vessel.
The HPCI system permits the nuclear plant to be shutdown, while maintaining sufficient reactor vessel water inventory until the reactor vessel is depressurized.
The HPCI system continues to operate until,reactor vessel pressure is below the pressure at which Low Pressure Coolant Injection (LPCI) or core spray system operation maintains core cooling.
The RCIC system provides makeup water to the reactor vessel during shutdown and isolation and following certain pipe break accidents to prevent the excessive release of radioactive materials to the environs as a result of inadequate core cooling.
A pipe break in the HPCI or RCIC supply piping results in a loss of reactor coolant inventory as well as a breach of the reactor coolant pressure boundary with a subsequent release path for fission products.
A high temperature in the vicinity of the break will occur owing to the release of energy from the reactor coolant pressure boundary in the form of steam.
The HPCI and RCIC systems both have four sets of four bimetallic temperature switches located in areas along the path of the steam supply piping of each system.,
The sixteen temperature switches are arranged into two divisional trip systems with eight temperature switches in each.
Each divisional trip logic scheme for the temperature switches is arranged in a one-out-of-two taken twice logic configuration for each of the fou'r reactor building areas being monitored.
(RCIC pump room, HPCI pump room, RCIC piping in torus area, HPCI piping in torus area).
The safety basis for the high temperature trip settings utilized by the HPCI and RCIC steam line space high temperature isolation trip channels is to ensure that automatic closure of each system's primary containment isolation valves occurs to prevent the excessive loss of reactor coolant and the release of significant amounts of radioactive material from the nuclear system process barrier.
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Enclosure 3
Page 2 of 2
TVA has utilized computer modeling techniques to predict the temperature response of various reactor building zones to HELBS.
The results of the modeling indicate that temperatures below 200' may be present for various HPCI and RCIC line break scenarios.
Therefore, the current 200' value in the unit 2 technical specifications must be revised.
The HELB calculation concluded that the HPCI and RCIC steam line space high temperature trip setting must be lowered to ensure that the safety basis is not violated.
The upper analytical limits were determined and used as inputs to a calculation which determined the trip setpoints and allowable values for the temperature switches based on known inaccuracies associated with them.
The allowance for instrument inaccuracies in determining the actual trip setpoint provides conservative assurance that the trip function will be performed at or before reaching the analytical input values used by the computer simulation code to determine the temperature and pressure response of the reactor building.
No physical change to plant equipment was involved.
The new allowable values for the temperature switches will be placed in Table 3.2.B.
Trip settings will be established in plant instructions to ensure that the allowable values are not exceeded taking into account instrument drif-t and inaccuracies.
The instrument descriptions are being broken down by specific area to ensure that differing allowable values are specified.
The minimum number of channels operable per trip system is being revised to conform with the system design (figure 1).
The HPCI and RCIC instrument channels are similarly configured.
Each consists of two trip systems which contain.two channels apiece.
A channel consists of one temperature switch located in the pump room and three temperature switches in the torus area.
The pump room switches have a higher allowable value because that space is more confined than the torus area.
Note 1.E is being added to the notes for table inoperable channel to be fixed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> position.
If the inoperable channel cannot be out-of-service time.the channel must be placed performs the intended function of the channel.
HPCI and RCIC temperature switches.
Note 4 to and the descriptive information on temperature bases section 3.2.
3.2.B.
This note requires an or placed in the tripped restored in the allowable in the tripped condition which Note 1.E is referenced for the table 3.2.B is being deleted switch configuration placed in Table 4.2.B is also being revised to break the instrument descriptions down by specific location.
Bases section 3.2 is being revised to be consistent with the changes to Tables 3.2.B and 4.2.B.
The proposed changes to the technical specifications help ensure that for a postulated HPCI or RCIC steam line break, the system's primary containment isolation valves are closed to prevent excessive loss of reactor coolant and the release of significant amounts of radioactive material from the nuclear system process barrier.
The trip settings have been selected high enough to prevent spurious isolation due to normal temperature excursions in the vicinity of the steam supply piping.
The changes are justified because they are conservative, do not increase the incidence of spurious trips, and provide a documented basis for the HPCI and RCIC steam line space high temperature trip settings.
HPCI/RCIC TEMPERATURE SWITCH LAYOUT TYPICALFOR HPCI OR RCIC SYSTEM TEMPERATURE SWZTCHES TRIP SYSTEM1 I TRIP SYSTEM2 TEMPERATVAE SWZTCHES PUMP AOOM TOAVS AREA 1 TORUS AREAR TORUS AAEA3 PUMP AOOM TOAVS TOAVS TOAVS AREA 1 AREAR AREA 3 TAZP CHANNELAELAY1 TAZP CHANNELAELhY 3 TEMPERATURE SWZTCHES TEMPERATURE SWZTCHES PUMP ROOM TOAVS AAEA 1 TORUS AREAR TORUS AAEA3 TRZP CHANNELAELAYZ I
I I
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PUMP AOOM TOAVS hAEA 1 TAZP CHANNELAELhY0 TORUS hAEA 1 TOAUS AAEA3 ZSOLATES HPCZ/ACZC FIGURE 1
ENCLOSURE 4 PROPOSED DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATION BROWNS FERRY NUCLEAR PLANT (BFN)
Descri tion of Pro osed Technical S ecification Chan e
The unit 2 technical specifications are being revised as follows.
1.
Table 3.2.B is being revised to incorporate new allowable values for the High Pressure Coolant Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) steam line space instrument channels and to indicate the specific areas (torus and pump rooms) where the temperature is being monitored.
A new note 1.E is being added to table 3.2.B and note 4 is being deleted.
2.
Table 4.2.B is being revised to indicate the specific areas (torus and pump rooms) where the HPCI and RCIC steam line space temperatures are being monitored.
3.
Bases section 3.2 is being revised to be consistent with the other changes.
Basis for Pro osed No Si nificant Hazards Consideration Determination 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 consideration 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 any accident previously evalua'ted, or 3) involve a significant reduction in a margin of safety.
1.
The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
The HPCI and RCIC steam line space high temperature isolations are provided to ensure automatic closure of each system's primary containment isolation valves for a HPCI or RCIC steam line break.
The isolation occurs when a very small leak has occurred.
If the small leak is allowed to continue without isolation, offsite dose limits may be reached.
TVA has utilized computer modeling techniques to predict the temperature response of various reactor building zones to high energy line breaks.
The results indicate that temperatures below the present 200' technical specification value may be present for various HPCI and RCIC line break scenarios.
The proposed change is being made so that HPCI and RCIC steam line breaks will be detected and isolated at the same or lower area temperatures.
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Enclosure 4
Page 2 of 2
The change to the allowable values of temperature in Table 3.2.B is in a conservative direction and provides the same or earlier detection and isolation of HPCI and RCIC steam line breaks.
A new note 1.E is added to Table 3.2.B and referenced for the HPCI and RCIC high temperature instrument channels which requires an inoperable channel to be tripped.
Both Tables 3.2.B and 4.2.B are revised to differentiate between temperature monitoring areas with different allowable values.
Changes to the bases are being made so it is consistent with the tables.
The changes have no effect on the probability of an accident and they will reduce the consequences of an accident through the same or earlier detection and isolation.
The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
The proposed change to the HPCI/RCIC steam line space high temperature isolations does not involve any modification to plant equipment.
No new failure modes are introduced.
There is no effect on the function or operation of any other plant system.
No new system interactions have been introduced by the change.
The results of a break in the HPCI or RCIC steam lines remain as before.
The HPCI or RCIC steam line area temperature switches will still detect a break due to an increase in area temperatures and close the system primary containment isolation valves to prevent reactor coolant loss.
The proposed change will conservatively serve to detect and mitigate HPCI and RCIC line breaks more expeditiously.
The proposed change does not involve a significant reduction in a margin of safety.
The margin of safety will be enhanced by ensuring that HPCI and RCIC steam line breaks are isolated at the same or lower steam line area temperatures.
Computer modeling techniques were utilized to predict
-the temperature response in various areas through which the HPCI and RCIC-steam lines pass.
The setpoints are established above the maximum expected room temperatures to avoid spurious actions due to ambient conditions and below the analytical limits to ensure timely pipe break detection and isolation.
The design and function of the affected components has not been changed.
Hinimum No.
Operable Per
~Tri S
s 1
1(2)
Function HPCI Trip System bus power monitor RCIC Trip System bus power monitor Instrument Channel Condensate Header Low Level (LS-73-55A 4 8)
TABLE 3 '.B (Continued)
Tri L v 1
tin N/A N/A
) Elev.
551'ction Remark 1.
Noni tors avail abil ity of power to logic systems.
1.
Honi tors avail abi1 ity of power to logic, systems.
1.
Below trip setting will open HPCI suction valves to the suppression chamber.
1(2) 2(2) 3(2) 3(2)
Instrument Channel Suppression Chamber High Level Instrument Channel-R'eactor High Mater Level (LIS-3-208A and LIS-3-208C)
Instrument Channel-RCIC Turbine Steam Line High Flow (PDIS-71-lA and 18)
Instrument Channel-RCIC Steam Supply Pressure - Low (PS 71-1A-D)
Instrument Channel-RCIC Turbine Exhaust Diaphragm Pressure High (PS 71-11A-D)
< 7" above instrument zero A
< 583" above vessel zero
< 450" H20 (7)
>50 psig
<20 psig l.
Above trip setting will open HPCI suction valves to the suppression chamber.
l.
Above trip setting trips RCIC turbine.
l.
Above trip setting isolates RCIC system and trips RCIC turbine.
l.
Below trip setting isolates
'RCIC system and trips RCIC turbine.
1.
Above trip setting isolates RCIC system and trips RCIC turbine.
TABLE 3'.B (Continued)
Hinimum Ho.
Operable Per
~Tri S
1 2(2)
Fun ion Instrument Channel Reactor High Water Level (LIS-3-2088 and LIS-3-208D)
Tri Level S t in
<583" above vessel zero.
~ACti c A
R mark l.
Above trip setting trips HPCI turbine.
3(2)
Instrument Channel
<90 psi (7)
HPCI Turbine Steam Line High Flow (PDIS-73-lA and 1B)
Instrument Channel-
>100 psvg HPCI Steam Supply Pressure - Low (PS 73-lA-D) l.
Above trip setting isolates HPCI system and trips HPCI turbine.
1.
Below trip setting isolates HPCI system and trips HPCI turbine.
3(2) 1 (16)
Instrument Channel-HPCI Turbine Exhaust Diaphragm (PS 73-20A-D)
Core Spray System Logic'CIC System (Initiating)
Logic RCIC System (Isolation)
Logic ADS Logic RHR (LPCI) System (Initiati on)
<20 psig N/A 0/A N/A N/A
'N/A l.
Above trip setting isolates HPCI system and trips HPCI turbine.
l.
Includes testing auto initiation inhibit to Core Spray Systems in other units.
l.
Includes Group 7 valves.
Refer to Table 3.7.A for list of valves.
1.
Includes Group 5 valves.
Refer to Table 3.7.A for list of valves.
TABLE 3.2.8 (Continued)
Kinimum No.
Operable Per
~Tri ~l 1(10) 1(10) 1(10) 1(11) 1(11) 1(12) 1(12) 1(13)
Func i n
Instrument Channel Thermostat (Core Spray Area Cooler Fan)
RHR Area Cooler Fan Logic Core Spray Area Cooler Fan Logic Instrument Channel Core Spray Kotors A or D Start Instrument Channel-Core Spray Kotor 8 or C Start
. Instrument Channel Core Spray Loop 1 Accident Signal (15)
Instrument Channel Core Spray Loop 2 Accident Signal (15)
RHRSW Initiate Logic RPT Logic Tri L v 1
e in
< 1004F N/A N/A N/A N/A N/A
~
N/A.
N/A N/A
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(17)
R m rk l.
Above trip setting starts Core Spray area cooler fans.
1.
Starts RHRSW pumps Al, 63, Cl, and D3 1.
Starts RHRSW pumps Al, B3, Cl, and 03 1.
Starts RHRSW pumps Al, 83, Cl, and D3 1.
Starts RHRSW pumps Al, B3, Cl, and 03 1.
Trips recirculation pumps on turbine control valve fast closure or stop valve closure
> 30% power.
TABLE 3.2.B (Continued)
Hinimum No.
Operable Per
~Tri ~LsS1 1(16) 1(16)
Fun i n ADS Timer ADS High Drywell Pressure Bypass Timer RCIC Steam Line Space Torus Area High Temperature RCIC Steam Line Space RCIC Pump Room Area High Temperature Allowable V 1 t<115 sec.
t<322 sec.
<1550 F
<1800 F
~A'on Rmrk l.
Above trip setting in conjunction with low reactor, water level permissive, low reactor water level;high
, drywell pressure or ADS high drywell pressure bypass timer timed out, and RHR or CSS pumps running, initiates ADS.
l.
Above trip setting, in conjuntion with low reactor water level permissive, low reactor water level, ADS timer timed out and RHR or CSS pumps running, initiates ADS.
l.
Above trip setting isolates RCIC system and trips RCIC turbine.
l.
Above trip setting isolates RCIC system and trips RCIC turbine.
HPCI Steam Line Space Torus Area
~
High Temperature
<180' l.
Above trip setting isolates HPCI system and trips HPCI turbine.
HPCI Steam Line Space HPCI Pump Room Area High Temperature
<200' l.
Above trip setting isolates HPCI system and trips HPCI turbine.
NOTES FOR T BLE 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.
E.
Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> restore the inoperable channel(s) to OPERABLE status or place the inoperable channel(s) in the tripped condition.
2.
In only one trip system.
3.
Not considered in a trip system.
4.
Deleted.
5.
With diesel power, each RHRS pump is scheduled to start immediately and each CSS pump is sequenced to start about 7 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 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
OTES FOR TAB E B
(Cont'd) 10.
Only one trip system for each cooler fan.
ll. In only two of the four 4160-V shutdown boards.
See note 13.
12.
In only one of the four '4160-V shutdown boards.
See note 13.
13.
An emergency 4160-V shutdown board is considered a trip system.
14.
RHRSW pump would be inoperable.
Refer to Section 4.5.C for the requirements of a RHRSW pump being inoperable.
15.
The accident signal is the satisfactory completion of a one-out-of-two taken twice logic of the drywell high pressure plus low reactor pressure or the vessel low water level (g 378" above vessel zero) originating in the core spray system trip system.
16.
The ADS circuitry is capable of accomplishing its protective action with one OPERABLE trip system.
Therefore, one trip system may be taken out of service for functional. testing and calibration for a period not to exceed eight hours.
17.
Two RPT systems exist, either of which will trip both recirculation pumps.
The systems will be individually functionally tested monthly. If the test period for one RPT system exceeds two consecutive
- hours, the system will be declared inoperable.
If both RPT systems are inoperable or if one RPT system is inoperable for more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, an orderly power reduction shall be initiated and reactor power shall be less than 30 percent within four hours.
18.
Not required to be OPERABLE in the COLD SHUTDOWN CONDITION.
BFN Unit 2 3.2/4.2-24
TABLE 4.2.8 SURVEILLANCE RE()UIREHENTS FOR INSTRUHENTATION THAT INITIATE OR CONTROL THE CSCS Functi n
Instrument Channel Reactor Low Water Level (LIS-3-58A-D)
Instrument Channel Reactor Low Water Level (LIS-3-184 & 185)
Instrument Channel Reactor Low Water Level (LIS-3-52 & 62A)
Instrument Channel Drywell High Pressure (PIS-64-58E-H)
Instrument Channel Drywell High Pressure (PIS-64-58A-D)
Instrument Channel Drywell High Pressure (PIS-64-57A-D)
F n i nal T
(1) (27)
(1) (27)
(1) (27)
(1) (27)
(1) (27)
(1) (27) libr i n
Once/18 Honths (28)
Once/18 Honths (28)
Once/18 Honths (28)
Once/18 Honths (28)
Once/18 Honths (28)
Once/18 Honths (28)
In rumen Ch c Once/day Once/day Once/day none none none Instrument Channel Reactor Low Pressure (PIS-3-74A&B, PS-3-74A&B)
(PIS-68-95, PS-68-95)
(PIS-68-96, PS-68-96)
(1) (27)
Once/6 Honths (28) none CO l' 4 lc
TABLE 4.2.B (Continued)'URVEILLANCE REQUIREHENTS FOR INSTRUHENTATION THAT INITIATE OR CONTROL THE CSCS Fn in Core Spray Auto Sequencing Timers (Normal Power)
Core Spray Auto Sequencing Timers (Diesel Power)
LPCI Auto Sequencing Timers (Normal Power)
LPCI Auto Sequencing Timers (Diesel Power)
RHRSW Al, B3, Cl, D3 Timers (Normal Power)
- RHRSW Al, 83, Cl, 03 Timers (Diesel Power)
I ADS Timer ADS High Drywell Pressure Bypass Timer RCIC Steam Line Space Torus Area High Temperature RCIC Steam Line Space RCIC Pump Room Area High Temperature n 'nlT (4)
(4)
(4)
(4)
(4)
(4)
(4)
(4) libr i n Once/operating cycle Once/operating cycl e Once/operating cycle Once/operating cycle Once/operating cycl e Once/operating cycle Once/operating cycl e Once/operating cycle Once/3 months Once/3 months In rmn h
none none none none none none none none none none
TABLE 4.2.B (Continued)
SURVEILLANCE RE(UIREHENTS FOR INSTRUHENTATION THAT INITIATE OR CONTROL THE CSCS Fun Fn inl T ibr i
n Instrument Channel-RHR Pump Discharge Pressure Instrument Channel-Core Spray Pump Discharge Pressure Once/3 months Once/3 months none none Core Spray Sparger to RPV d/p Trip System Bus Power Honitor Instrument Channel Condensate Header Low Level (LS-73-56A, 8)
Instrument Channel Suppression Chamber High Level Instrument Channel Reactor High Water Level Instrument Channel.-
RCIC Turbine Steam Line High Flow Instrument Channel-RCIC Steam Supply Low Pressure Instrument Channel-RCIC Turbine Exhaust Diaphragm High Pressure HPCI Steam Line Space Torus Area High Temperature HPCI Steam Line Space HPCI Pump Room Area High Temperature Once/operating Cycle Once/31 days Once/31 days Once/3 months N/A Once/3 months Once/3 months Once/3 months Once/3 months Once/18 months Once/18 months Once/3 months Once/3 months Once/day none none none Once/day none none none none none
TABLE 4.2.8 (Continued)
SURVEILLANCE REQUIREHENTS FOR INSTRUHENTATION THAT INITIATE OR CONTROL THE CSCS F n Instrument Channel-HPCI Turbine Steam Line High Flow Instrument Channel-HPCI Steam Supply Low Pressure Instrument Channel-HPCI Turbine Exhaust Diaphragm High Pressure Core Spray System Logic RCIC System (Initiating) Logic RCIC System (Is ol ati on) Logic HPCI Sys'em (Initiating) Logic HPCI System (Isolati'on) Logic ADS Logic
'PCI (Initiating) Logic LPCI (Containment Spray)'ogic Core Spray System Auto Initiation Inhibit (Core Spray Auto Initiati on)
LPCI Auto Initiation Inhibit (LPCI Auto Initiation) n nl T
Once/31 days Once/31 days Once/18 months Once/18 months Once/18 months Once/18 months
'nce/18 months Once/18 months Once/18 months Once/18 months Once/18 months (7)
Once/18 months (7) 1'br i n Once/3 months Once/18 months Once/18 months (6)
N/A (6)
(6)
(6)
(6)
(6)
N/A N/A In rum n h
none none none N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A
3.2 BASES (Cont'd) flow instrumentation is a backup to the temperature instrumentation.
In the event of a loss of the reactor building ventilation system, radiant heating in the vicinity of the main steam lines raises the ambient temperature above 200 F.
The temperature increases can cause an unnecessary main steam line isolation and reactor scram.
Permission is provided to bypass the temperature trip for four hours to avoid an unnecessary plant transient and allow performance of the secondary containment leak. rate test or make repairs necessary to regain normal ventilation.
High radiation monitors in the main steam line tunnel have been provided to detect gross fuel failure as in the control rod drop accident.
With the established nominal setting of three times normal background and main steam line isolation valve closure, fission product release is limited so that 10 CFR 100 guidelines are not exceeded for this accident.
Reference Section 14.6.2 FSAR.
An alarm with a nominal setpoint of 1.5 x normal full-power background is provided also.
Pressure instrumentation is provided to close the main steam isolation valves in RUN Mode when the main steam line pressure drops below 825 psig.
The HPCI high flow and temperature instrumentation are provided to detect a break in the HPCI steam piping.
Tripping of this instrumentation results in actuation of HPCI isolation valves.
Tripping logic for the high flow is a 1-out-of-2 logic, and all sensors are required to be OPERABLE.
High temperature in the vicinity of the HPCI equipment is sensed by four sets of four bimetallic temperature switches.
The 16 temperature switches are arranged in two trip systems with eight temperature switches in each trip system.
Each trip system consists of two elements.
Each channel contains one temperature switch located in the pump room and three temperature switches located in the torus area.
The RCIC high flow and hi,gh area temperature sensing instrument channels are arranged in the same manner as the HPCI system.
The HPCI high steam flow trip setting of 90 psid and the RCIC high steam flow trip setting of 450" H20 have been selected such that the trip setting is high enough to prevent spurious tripping during pump startup but low enough to prevent core uncovery and maintain fission product releases within 10 CFR 100 limits.
The HPCI and RCIC steam line space temperature switch trip settings are high enough to prevent spurious isolation due to normal temperature excursions in the vicinity of the steam supply piping.
Additionally, these trip settings ensure that the primary containment isolation steam supply valves isolate a break within an acceptable=time period to prevent core uncovery and maintain fission product releases within 10 CFR 100 limits.
High temperature at the Reactor Water Cleanup (RWCU) System floor drain in the space near. the RWCU system or'in the space near the pipe trench containing RWCU piping could indicate a break in the cleanup system.
When high temperature
- occurs, the cleanup system is isolated.
'nit 2 3.2/4.2-67
f'
3.2 BASES (Cont'd)
The instrumentation which initiates CSCS action is arranged in a dual bus system.
As for other vital instrumentation arranged in this fashion, the specification preserves the effectiveness of the system even during periods when maintenance or testing is being performed.
An exception to this is when logic functional testing is being performed.
The control rod block, functions are provided to prevent excessive control rod withdrawal so that MCPR does not decrease to 1.07.
The trip logic for this function is 1-out-of-n:
e.g.,
any trip on one of six APRMs, eight IRMs, or four SRMs will result in a rod block.
The minimum instrument channel requirements assure sufficient instrumentation to assure the single failure criteria is met.
The minimum instrument channel requirements for the RBM may be reduced by one for maintenance,
- testing, or calibration.
This does not significantly increase the risk of an inadvertent control rod withdrawal, as the other channel is available, and the RBM is a backup system to the written sequence for withdrawal of control rods.
The APRM rod block function is flow biased and prevents a significant reduction in MCPR, especially 'during operation at reduced flow.
The APRM provides gross core protection; i.e., limits the gross core power increase from withdrawal of control rods in the normal withdrawal sequence.
The trips are set so that MCPR is maintained greater than 1.07.
The RBM rod block function provides local protection of the core; i.e.,
the prevention of criti'cal power in a local region of the.core, for a single rod withdrawal error from a limiting control rod pattern.
If the IRM channels are in the worst condition of allowed bypass, the sealing arrangement is such that for unbypassed IRM channels, a rod block signal is generated before the detected neutrons flux has increased by more than a factor of 10.
A downscale indication is an indication the instrument has failed or the instrument is not sensitive enough.
In either case the instrument will not respond to changes in control rod motion and thus, control rod motion is prevented.
The refueling interlocks also operate one logic channel, and are required for safety only when the mode switch is in the refueling position.
For effective emergency core cooling for small pipe breaks, the HPCI system must function since reactor, pressure does not decrease rapid enough to allow either core spray or LPCI to operate in time.
The automatic pressure relief function is provided as a backup to the HPCI in the event the HPCI does not operate.
The arrangement of the tripping contacts is such as to provide this function when necessary and minimize spurious operation.
The trip settings given in the. specification are BFN Unit 2 3.2/4.2-68
3.2 BASES (Cont'd) adequate to assure the above criteria are met.
The specification preserves the effectiveness of the system during periods of maintenance, testing, or calibration, and also minimizes the risk of inadvertent operation; i.e., only one instrument channel out of service.
Two post treatment offgas radiation monitors are provided and, when their trip point is reached, cause an isolation of the offgas line.
Isolation is initiated when both instruments reach their high trip point or one has an upscale trip and the other a downscale trip or both have a downscale trip.
Both instruments are required for trip but. the instruments.are set so that the instantaneous stack release rate limit given in Specification 3.8 is not exceeded.
Four radiation monitors are provided for each unit which initiate Primary Containment Isolation (Group 6 isolation valves) Reactor Building Isolation and operation of the Standby Gas Treatment System.
These instrument channels monitor the radiation in the reactor zone ventilation exhaust ducts and in the refueling zone.
C Trip setting of 100 mr/hr for the monitors in the refueling zone are based upon initiating normal ventilation isolation and SGTS operation so that none of the activity released during the refueling accident leaves the Reactor Building via the normal ventilation path but rather-all the activity is processed by the SGTS.
Flow integrators and sump fillrate and pump out rate timers are used to determine leakage in the drywell.
A system whereby the time interval t'o filla known volume will be utilized to provide a backup.
An air sampling system is also provided to detect leakage inside the primary containment (See Table 3.2.E).
For each parameter monitored, as listed in Table 3.2.F, there are two channels of instrumentation except as noted.
By comparing readings between the two channels, a near continuous surveillance of instrument performance is available.
Any deviation in readings will initiate an early recalibration, thereby maintaining the quality of the instrument readings.
Instrumentation is provided for isolating the control room and initiating a pressurizing system that processes outside air before supplying it to the control room.
An accident signal that isolates primary'containment will also automatically isolate the control room and initiate the emergency pressurization system.
In addition, there are radiation monitors in the normal ventilation system that will isolate the control room and initiate the emergency pressurization system.- Activity required to cause automatic actuation is about one mRem/hr.
Because of the constant surveillance and control exercised by TVA over the Tennessee Valley, flood levels of large magnitudes can be predicted in BFN Unit 2 3.2/4.2-69
4P
3.2 BASES (Cont'd) advance of their actual, occurrence.
In all cases, full advantage will be taken of advance warning to take appropriate action whenever reservoir levels above normal pool are predicted;
- however, the plant flood protection is always in place and does not depend in any way on advanced warning.
Therefore, during flood conditionsp the plant will be permitted to operate until water begins to run across the top of the pumping station at elevation 565.
Seismically qualified, redundant level switches each powered from a separate division of power are provided at the pumping station to give main control room indication of this condition.
At that time. an orderly shutdown of the plant will be initiated, although surges even to a depth of several feet over the pumping station deck will not cause the loss of.the main condenser circulating water pumps.
The operability of the meteorological instrumentation ensures that sufficient meteorological data is available for estimating potential radiation dose to the public as a result of routine or accidental release of radioactive materials to the atmosphere.
This capability is required to evaluate the need for initiating protective measures to protect the health and safety of the public.
The operability of the seismic instrumentation ensures that sufficient capability is available to promptly determine the seismic response of those features important to safety.
This capabi.lity is required to permit comparison of the measured response to that used in the design basis for Browns Ferry Huclear Plant and to determine whether the plant can continue to be operated safely.
The instrumentation provided is consistent with specific portions of the recommendations of Regulatory Guide 1.12 "Instrumentation for Earthquakes."
The ra'dioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous effluents.
The alarm/trip setpoints for these instruments will be calculated in accordance with guidance provided in the ODCM to ensure that the alarm/trip will occur prior to exceeding the limits of 10 CFR Part 20.
This instrumentation also includes provisions for monitoring the concentration of potentially explosive gas mixtures in the offgas holdup system.
The operability and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50; The radioactive liquid effluent instrumentation is provided to monitor and
- control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases of liquid effluents.
The alarm/trip setpoints for these instruments shall be calculated in accordance with guidance provided in the ODCM to ensure that the alarm/trip will occur prior to exceeding the limits of 10 CFR Part 20 Appendix B, Table II, Column 2.
The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50.
BFN Unit 2 3.2/4.2'-70
3.2 BASES (Cont'd)
ATWS/RPT, Anticipated Transients without Scram/Recirculation Pump Trip system provides a means of limiting the consequences of the unlikely occurrence of a fai,lure to scram during an ATWS event.
The response of the plant to this postulated event (ATWS/RPT) follows the BWR Owners Group Report by General Electric HEDE-31096-P-A and the accompanying HRC Staff Safety Evaluation Report.
ATWS/RPT utilizes the engineered safety feature (ESF) master/slave analog trip units (ATU) which consists of four level and four pressure channels total.
The initiating logic consists of two independent trip systems each consisting of two reactor dome high pressure channels and two reactor vessel low level channels.
A coincident trip of either two low levels or two high pressures in the same trip system causes initiation of ATWS/RPT.
This signal from either trip system opens one of two EOC (end-of-cycle) breakers in series (the other system opens the other breaker) between the pump motor and the Motor Generator set driving each recirculation pump.
Both systems are completely redundant such that only one trip system is necessary to perform the ATWS/RPT function.
Power comes from the 250 VDC shutdown boards.
Setpoints for reactor dome high pressure and reactor vessel low level are such that a normal Reactor Protection System scram and accompanying recirculation pump trip would occur before or coincident with the trip by ATWS/RPT.
4.2 BASES The instrumentation listed in Tables 4.2.A through 4.2.F will be functionally tested and calibrated at regularly scheduled intervals.
The same design reliability goal as the Reactor Protection System of 0.99999 generally applies for all applications of (1-out-of-2) X (2) logic.
Therefore, on-off sensors are tested once/3 months, and bistable trips associated with analog sensors and amplifiers are tested once/week.
Those instruments which, when tripped, result in a rod block have their contacts arranged in a 1-out-of-n logic, and all are capable of being bypassed.
For such a tripping arrangement with bypass capability
- provided, there is an optimum test interval that should be maintained in order to maximize the reliability of a given channel (7).
This takes account. of the fact that testing degrades reliability and the optimum interval between tests is approximately given by:
Where:
the optimum interval between tests.
the time the trip contacts are disabled from performing their function while the test is in progress.
the expected failure rate of the relays.
BFH Unit 2 3.2/4.2-71
~
4.2 /USES (Cont'd)
To test the trip relays requires that the channel be bypassed, the test
- made, and the system returned to its initial state.
It is assumed this task requires an estimated 30 minutes to complete in a thorough and workmanlike manner and that the relays have a failure rate of 10 6
failures per hour.
Using,this data and the above operation, the optimum test interval is:
= 1 x 10'6 10
= 40 days or additional ar in a test interval of once er month will be used initiall The sensors and electronic apparatus have not been included here as these are analog devices with readouts in the control room and the sensors and electronic apparatus can be. checked by comparison with other. like instruments.
The checks which are made on a daily basis are adequate to assure operability of the sensors and electronic apparatus, and the test interval given above provides for optimum testing of the relay circuits.
The above calculated test interval optimizes each individual channel, considering it to be independent of all others.
As an example, assume that there are two channels with an individual technician assigned to each.
Each technician tests his channel at the optimum frequency, but the two technicians are not allowed to communicate so that one can advise the other that his channel is under test.
Under. these conditions, it is possible for both channels to be under test simultaneously.
Now, assume that the technicians are required to communicate and that two channels are never tested at the same time.
(7) UCRL-50451, Improving Availability and Readiness of Field Equipment Through Periodic Inspection, Benjamin Epstein, Albert Shiff, July 16,
- 1968, page 10, Equation (24), Lawrence Radiation Laboratory.
Forbidding simultaneous testing improves the availability of the system over that which would be achieved by testing each channel independently.
These one-out-of-n trip systems will be tested one at a time in order to take advantage of this inherent improvement in availability.
Optimizing each channel independently may not truly optimize the system considering the overall rules of system operation.
However, true system optimization is a complex problem.
The optimums are broad; not sharp, and optimizing the, individual channels is generally adequate for the system.
The formula given above minimizes the unavailability of a single channel which must be bypassed during testing.
The minimization of the unavailability is illustrated by Curve No.
1 of Figure 4.2-1 which assumes that a channel has a failure rate of O.l x 10 6/hour and 0.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> is required to test it.
The unavailability is a'minimum at a test interval i, of 3.16 x 103 hours0.00119 days <br />0.0286 hours <br />1.703042e-4 weeks <br />3.91915e-5 months <br />.
BFN Unit 2 3.2/4.2-72
t" t
4,
4.2 'BASES (Cont'd)
If two similar channels are used in a 1-out-of-2 configuration, the test interval for minimum unavailability changes as a function of the rules for testing.
The simplest case is to test each one independent of the other.
In this 'case, there is assumed to be a finite probability that both may be bypassed at one time.
This case is shown by Curve Ho. 2.
Note that the unavailability is lower as expected for a redundant system and the minimum occurs at the same test interval.
Thus, if the two channels are tested independently, the equation above yields the test interval for minimum unavailability.
A more usual case is that the testing is not done independently.
If both channels are bypassed and tested at the same time, the result is shown in Curve No. 3.
Note that the minimum occurs at about 40,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />, much longer than for cases 1 and 2.
Also, the minimum is not nearly as low as Case 2 which indicates that this method of testing does not take full advantage of the redundant channel.
Bypassing both channels for simultaneous testing should be avoided.
The most likely case would be to stipulate that one channel be bypassed,
- tested, and restored, and then immediately following, the second channel be bypassed,
- tested, and restored.
This is shown by Curve Ho. 4.
Note that there is no true minimum.
The curve does have a definite knee and very little reduction in system unavailability is achieved by testing at a shorter interval than computed by the equation for a single channel.
The best test procedure of all those examined is to per'fectly stagger the tests.
That is, if the test interval is four months, test one or the other channel every two months.
This is shown in Curve No. 5.
The difference between Cases 4 and 5 is negligible.
There may be other arguments, however, that more strongly support the perfectly staggered
- tests, including reductions in human error.
The conclusions to be drawn are these:
1.
A 1-out-of-n system may be treated the same as a single channel in terms of choosing a test interval; and 2.
more than one channel should not be bypassed for testing at any one time.
The radiation monitors in the refueling area ventilation duct which initiate building isolation and standby gas treatment operation are arranged in two 1-out-of-2 logic systems.
The bases given for the rod blocks apply here also and were used to arrive at. the functional testing frequency.
The offgas post treatment monitors are connected in a 2-out-of-2 logic arrangement.
Based on experience with instruments of similar design, a testing interval of once every three months has been found adequate.
The automatic pressure relief instrumentation can be considered to be a 1-out-of-2 logic system and the discussion above applies also.
BFN Unit 2 3.2/4.2-73
t
\\
I P
C
4.2
~B SES (Cont'd)
The criteria for ensuring the reliability and accuracy of the radioactive gaseous effluent instrumentation is list'ed in Table 4.2.K.
The. criteria for ensuring the reliability and accuracy of the radioactive liquid effluent instrumentaiton is listed in Table 4.2.D.
BFN Unit 2 3.2/4.2-73a
'gk fp 4b
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