ML16071A150: Difference between revisions

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#REDIRECT [[NL-15-1898, Enclosure 4: EAL Verification and Validation Documents - License Amendment Request for Changes to EAL Schemes to Adopt NEI 99-01 Rev. 6 and to Modify Radiation Monitors at Farley Nuclear Plant. Part 3 of 6]]
| number = ML16071A150
| issue date = 03/03/2016
| title = Enclosure 4: EAL Verification and Validation Documents - License Amendment Request for Changes to EAL Schemes to Adopt NEI 99-01 Rev. 6 and to Modify Radiation Monitors at Farley Nuclear Plant. Part 3 of 6
| author name =
| author affiliation = Southern Nuclear Operating Co, Inc
| addressee name =
| addressee affiliation = NRC/NRR
| docket = 05000321, 05000348, 05000364, 05000366, 05000424, 05000425
| license number =
| contact person =
| case reference number = NL-15-1898
| package number = ML16071A108
| document type = License-Application for Facility Operating License (Amend/Renewal) DKT 50
| page count = 44
}}
 
=Text=
{{#Wiki_filter:V6Page 1 of 6SOUTHERN NUCLEAR DOCUMENT TYPE:PLANT E.I. HATCH ANNUNCIATOR RESPONSE PROCEDURE (ARP) AGE 1 OF 2DOCUMENT TITLE: DOCUMENT NUMBER:/
VERSION NO:ARP'S FOR CONTROL PANEL 1 H11-P654, 34AR-654-901-1 2.ALARM PANEL 1 2.EXPIRATION APPROVALS:
EFFECTIVE DATE: DEPARTMENT MANAGER DAK for KDL DATE 09-07-12 DATE:09-07-2012 N/A SSM /PM N/A DATE N/AANNUNCIATOR RESPONSE PROCEDURES FOR 1H11-P654
-ALARM PANEL 1ARP NO.VER NO.6.1ARP NO.VER NO.ARP NO.VER. NO.SPARE34AR-654-001
-134AR-654-031-1 4.034AR-654-061 34AR-654-002-1 1.0 34AR-654-032 SPARE 34AR-654-062-1 3.034AR-654-003-1 4.1 34AR-654-033-1 SPARE 34AR-654-063-1 2.034AR-654-004-1 6.1 34AR-654-034-1 2.0 34AR-654-064-1 3.034AR-654-005-1 3.0 34AR-654-035-1 2.0 34AR-654-065-1 7.334AR-654-006-1 7.1 34AR-654-036-1 1.0 34AR-654-066-1 2.034AR-654-007-1 12.0 34AR-654-037-1 4.0 34AR-654-067-1 8.034AR-654-008-1 4.0 34AR-654-038-1 4.0 34AR-654-068-1 8.034AR-654-009-1 5.2 34AR-654-039-1 2.0 34AR-654-069-1 2.034AR-654-0 10-1 2.1 34AR-654-040-1 2.0 34AR-654-070-1 3.134AR-654-01 1 SPARE 34AR-654-041-1 2.1 34AR-654-071-1 2.034AR-654-012-1 0.0 34AR-654-042-1 1.0 34AR-654-072-1 2.034AR-654-01 3-1 2.0 34AR-654-043-1 2.0 34AR-654-073-1 2.034AR-654-014-1 3.0 34AR-654-044-1 3.0 34AR-654-074-1 3.034AR-654-015 SPARE 34AR-654-045 SPARE 34AR-654-075-1 SPARE34AR-654-0 16-1 4.0 34AR-654-046-1 3.0 34AR-654-076-1 3.034AR-654-017 SPARE 34AR-654-047-1 1.0 34AR-654-077-1 1.034AR-654-01 8-1 3.1 34AR-654-048-1 3.0 34AR-654-078-1 1.134AR-654-01 9-1 2.0 34AR-654-049-1 3.0 34AR-654-079-1 2.034AR-654-020-1 2.0 34AR-654-050-1 4.1 34AR-654-080-1 6.234AR-654-02 1-1 5.4 34AR-654-05 1-1 3.0 34AR-654-08 1-1 SPARE34AR-654-022-1 11.1 34AR-654-052
-SPARE 34AR-654-082 SPARE34AR-654-023-1 4.0 34AR-654-053
.SPARE 34AR-654-083 SPARE34AR-654-024-1 2.0 34AR-654-054-1 3.0 34AR-654-084-1 2.034AR-654-025 SPARE 34AR-654-055-1 3.0 34AR-654-085-1 3.034AR-654-026 SPARE 34AR-654-056-1 1.1 34AR-654-086-1 2.034AR-654-027 SPARE 34AR-654-057-1 1.0 34AR-654-087-1 2.034AR-654-028-1 1.0 34AR-654-058-1 2.0 34AR-654-088-1 2.034AR-654-029-1 2.0 34AR-654-059-1 3.0 34AR-654-089-1 3.034AR-654-030 SPARE 34AR-654-060-1 0.0 34AR-654-090-1 SPARENOTE. Approval signature on this page constitutes approval for all procedures listed above at the!version indicated.
Tab numbers in the back correspond to procedure sequence number. jILevel Of Use ARPsCONTINUOUS ALLREFERENCE NoneINFONoneNMP-AP-002 V6Page 2 of 6PAGE 2 OF2SOUTHERN NUCLEARPLANT E.I. HATCHDOCUMENT TITLE: DOCUMENT NUMBER:ARP'S FOR CONTROL PANEL 1HI11-P654,I 34AR-654-901-1 ALARM PANEL 1VERSION NO:23.0UNIT 11 H11-P654
-1 (Left)654-00 1 654-002 654-003 654-004 654-005 654-006 654-007 654-008654-016 654-017 654-018 654-019 654-020 654-021 654-022 654-023654-03 1 654-032 654-033 654-034 654-035 654-036 654-037 654-038654-046 654-047 654-048 654-049 654-050 654-05 1 654-052 654-053654-06 1 654-062 654-063 654-064 654-065 654-066 654-067 654-068654-076 654-077 654-078 654-079 654-080 654-08 1 654-082 654-083UNIT I1H11-P654
-1 (Right)654-009 654-010 654-011 654-012 654-013 654-014 654-015654-024 654-025 654-026 654-027 654-028 654-029 654-030654-039 654-040 654-04 1 654-042 654-043 654-044 654-045654-054 654-055 654-056 654-057 654-058 654-059 654-060654-069 654-070 654-07 1 654-072 654-073 654-074 654-075654-084 654-085 654-086 654-087 654-088 654-089 654-090NMP-AP-002 V6Page 3 of 61.0 IDENTIFICATION:
ALARM PANEL 654SPENT FUELSTORAGE POOLLEVEL LOWDEVICE:1G41-N362 Level Sensor1G41-N372 Remote Electronics SETPOINT:
225 ft. 9 in.(21' -7" above the top of the fuel assemblies seated inthe Fuel Pool)5.0 OPERATOR ACTIONS:* Water level shall be maintained at least 21' above the top of the upper tie plates ofthe irradiated fuel assemblies seated in the fuel storage racks.Normal water level is 22' -4" to 22'- 7", as indicated on 1T24-R001.
* Water may be added to the fuel pool from the following sources:* CST, via 1G41-F041 NOTES:* Service Water, via 1G41-F217
* Fire Protection hose station* The suction piping for the DHR system has anti-siphon holes at elev. 225' 7".If level continues to drop, air could be drawn into the suction and air bind or damagethe DHR pump.5.1 Enter 34AB-G41-002-1, Decreasing Rx Well/Fuel Pool Water Level. Li5.2 IF Fuel Pool gates are installed:
5.2.1 Raise water level by regulating 1G41-F041, Spent Fuel Pool Make-up water from CSTValve, located at 185RBR07, panel 1H21-P155.
LI5.2.2 Confirm Fuel Pool Cooling filter effluent is returning to fuel pool per 34SO-G41-003-1, Fuel Pool Cooling and Cleanup System. LI5.3 IF Fuel Pool gates are NOT installed, request Maintenance to install gates. LI5.4 If the DHR system is running with suction aligned to the Unit 1 Fuel Pool,then at the direction of the SS, secure the DHR system. LI6.0 CAUSES:6.1 Water loss from normal evaporation 6.3 Malfunction of level switch (fail safe)6.2 System leakage
 
==7.0 REFERENCES==
:
8.0 TECH. SPECS./TRMIODCM/FHA 7.1 H-16002, Fuel Pool Cooling System P&ID Unit One, Section 3.7.87.2 H-17074, Fuel Pool Cooling System G41 Elem Diag34AR-654-022-IVER. 11.1MGR-0048 Ver. 5.0NMPA02 NMP-AP-002 V6Page 4 of 6DOCUMNT TTLE:DOCUMENT NUMBER: VERSION NO:ARP'SALRFOR CONTROLpAE PANEL 2H11-P654, 34AR-654-901-2 23.4EXPI RATION APPROVALS:
EFFECTIVE DATE: DEPARTMENT MANAGER DAK for KDL DATE 09-07-12 DATE:2-27-15N/A SSM /PM N/A DATE N/A _____ANNUNCIATOR RESPONSE PROCEDURES FOR 2H11-P654
-ALARM PANEL 1ARP NO. VER. NO. ARP NO. VER. NO. ARP NO. VER. NO.34AR-654-001-2 6.4 34AR-654-031-2 3.3 34AR-654-061 SPARE34AR-654-002-2 2.1 34AR-654-032-2 2.1 34AR-654-062-2 3.134AR-654-003-2 3.1 34AR-654-033 SPARE 34AR-654-063 SPARE34AR-654-004-2 2.2 34AR-654-034-2 2.3 34AR-654-064-2 2.334AR-654-005-2 2.0 34AR-654-035-2 2.1 34AR-654-065-2 4.234AR-654-006-2 8.0 34AR-654-036-2 3.4 34AR-654-066-2 5.134AR-654-007-2 11.0 34AR-654-037-2 2.0 34AR-654-067-2 6.034AR-654-008-2 3.1 34AR-654-038-2 4.1 34AR-654-068-2 6.034AR-654-009 SPARE 34AR-654-039 SPARE 34AR-654-069 SPARE34AR-654-01 0 SPARE 34AR-654-040-2 3.0 34AR-654-070-2 4.234AR-654-01 1 SPARE 34AR-654-041-2 3.1 34AR-654-071-2 4.134AR-654-012-2 3.2 34AR-654-042-2 2.1 34AR-654-072-2 4.134AR-654-013-2 2.2 34AR-654-043-2 2.1 34AR-654-073-2 4.134AR-654-014-2 1.2 34AR-654-044-2 2.1 34AR-654-074-2 5.434AR-654-015-2 3.2 34AR-654-045 SPARE 34AR-654-075 SPARE34AR-654-016-2 4.3 34AR-654-046-2 3.1 34AR-654-076-2 2.1 __34AR-654-017-2 2.1 34AR-654-047-2 3.1 34AR-654-077-2 3.1 __34AR-654-018 SPARE 34AR-654-048 SPARE 34AR-654-078-2 3.1 __34AR-654-019-2 2.2 34AR-654-049-2 1.1 34AR-654-079-2 2.234AR-654-020-2 3.2 34AR-654-050 SPARE 34AR-654-080-2 7.534AR-654-021-2 7.1 34AR-654-051 SPARE 34AR-654-081 SPARE34AR-654-022-2 11.1 34AR-654-052 SPARE 34AR-654-082 SPARE34AR-654-023-2 6.1 34AR-654-053 SPARE 34AR-654-083 SPARE34AR-654-024 SPARE 34AR-654-054 SPARE 34AR-654-084 SPARE34AR-654-025 SPARE 34AR-654-055-2 3.0 34AR-654-085-2 4.234AR-654-026-2 4.1 34AR-654-056-2 3.1 34AR-654-086-2 4.1 __34AR-654-027-2 4.0 34AR-654-057-2 3.1 34AR-654-087-2 4.1 __34AR-654-028 SPARE 34AR-654-058 SPARE 34AR-654-088-2 4.134AR-654-029 SPARE 34AR-654-059-2 2.2 34AR-654-089 SPARE34AR-654-030 SPARE 34AR-654-060 SPARE 34AR-654-090 SPAREIIIINOTQE: I Approval signature on this page constitutes approval for all procedures listed above at theIversion indicated.
Tab numbers in the back correspond to procedure sequence number.Ilit rmLevel Of Use ARPsCONTINUOUS ALLREFERENCE NoneINFO None V6Page 5 of 62H11-P654-1 (Left)654-001 654-002 654-003 654-004 654-005 654-006 654-007 654-008654-016 654-017 654-018 654-019 654-020 654-021 654-022 654-023654-031 654-032 654-033 654-034 654-035 654-036 654-037 654-038654-046 654-047 654-048 654-049 654-050 654-051 654-052 654-053654-061 654-062 654-063 654-064 654-065 654-066 654-067 654-068654-076 654-077 654-078 654-079 654-080 654-081 654-082 654-083UNIT 22H11-P654
-1 (Right)654-009 654-010 654-011 654-012 654-013 654-014 654-015654-024 654-025 654-026 654-027 654-028 654-029 654-030654-039 654-040 654-041 654-042 654-043 654-044 654-045654-054 654-055 654-056 654-057 654-058 654-059 654-060654-069 654-070 654-071 654-072 654-073 654-074 654-075654-084 654-085 654-086 654-087 654-088 654-089 654-090 V6Page 6 of 61.0 IDENTIFICATION:
ALARM PANEL 654 ________" --t---SPENT FUELSTORAGE POOLLEVEL LOWDEVICE: SETPOINT:
2G41-N372 Remote Electronics 226' 2.5" (approx.
22' 0.5" above the top of2G41-N362 Level Sensor the seated fuel assemblies)
* Water level shall be maintained at least 21' above the top of the upper tie plates of the irradiated fuel assemblies seated in the fuel storage racks.iNormal water level is 22' -4" to 22' -7" as indicated on 2T24-R001.
* Water may be added to the fuel pool from the following sources:NOTES
* CST via 2G41-F054
* Service water via 2G41-F040
* Fire Protection hose station* The suction piping for the DHR system has anti-siphon holes at elev. 225' 7".If level continues to drop, air could be drawn into the suction and air bind or damage the DHR pump.5.1 Enter 34AB-G41-002-2, Decreasing Rx Well/Fuel Pool Water Level. Li5.2 IF Fuel Pool gates are installed:
5.2.1 Raise water level by regulating 2G41-F054, Spent Fuel Pool Make-up water from CST,located at 185RBR20, panel 2H21-P155.
LI5.2.2 Confirm the Fuel Pool Cooling Filter effluent is returning to Fuel Poolper 34S0-G41-003-2, Fuel Pool Cooling and Cleanup System. Li5.2.3 Confirm air supply pressure to seals is between 33 PSIG AND 37 PSIG AND2P51-F549, 2P51-F563, and 1P51-F555 (unit one's air supply valve), Air Supply Valves,are OPEN. Li5.3 IF Fuel Pool gates are NOT INSTALLED request maintenance to INSTALL gates. Li5.4 If the DHR system is running with suction aligned to the Unit 1 Fuel Pool,then at the direction of the SS, secure the DHR system. Li6.0 CAUSES:6.1 Water loss from normal evaporation 6.5 Reactor well leakage6.2 Fuel Pool liner leakage 6.6 Dryer/Separator storage pooi leakage6.3 Fuel Pool gate leakage 6.7 Fuel Pool to transfer canal gate leakage6.4 System leakage 6.8 Malfunction of level switch (fail safe)
 
==7.0 REFERENCES==
:
18.0 TECH. SPECS./TRM/ODCM/FHA:
7.1 H-26039, Fuel Pool Cooling System P&ID Tech Specs 3.7.8, Spent Fuel Storage7.2 H-27736, Fuel Pool Cooling System Elem Pool Water Level34AR-654-022-2 VER. 11.1 V7, Page 1 of 3Fission Product Barrier Emerqency Action LevelsFuel Clad Barrier:Emergency Action LevelsFuel Clad Barrier Potential Loss Threshold 2.ARPV water level cannot be restored and maintained above -155 inches or cannotbe determined.
(Ref H16 145 and H26189)According to COLR for HNP the currently used fuel is GE 14. According to NEDC-32868P Rev 5 Appendix A (Reference of the COLR) the fuel length for GEI4 fuelwas increased from 148" to 150" inches. The Appendix A is attached below. Thusthe top of the fuel per TS Bases 2.1.1.3 is 158.44 inches below instrument zero.According to 31EO-EOP-015-1 and 31EO-EOP-015-2 "CP1 Flow Chart"operators are instructed to maximize water injection rates from alternate injection subsystems when reactor water level drops below -155 inches of instrument zero. This value is more conservative than the actual TOAF level.
V7, Page 2of 3Fuel Clad Barrier Loss Threshold 4.ADWRRM greater than 1,400 R/hr.In Attachment C the detector radiation level of 1.4E3 R/hr was calculated.
Thecalculation used core inventory from NL-06-1637 to calculate isotopesconcentrations.
The calculation for DEI131 was performed to find a ratio to DEl300uCi/gm.
GRODEC was used to calculate the fluence within the dr'ywell.
Cylinder geometry was used to calculate the geometric fraction.
Fuel Clad Barrier Loss Threshold 5.AOffgas Pre- and Post-Treatment Monitors Offscale High AND Fission ProductMonitor Offscale High.Attachment D performed an evaluation for Offgas Pre- and Post-treatment monitors D1 1-K615 (section A of Attachment D) and Containment fission productmonitors D11IP010 (section B of Attachment D). It was found that theseinstruments will be off scale.RCS Barrier:Emergency Action LevelsRCS Barrier Loss Threshold I .APrimary containment pressure greater than 1.85 psig due to RCS leakage.LIS 1C71N650A-D 1.85 psigLIS 2C71N650A-D 1.85 psigReferences (H 16568, PDMS)Ret PDMSRef. PDMSRCS Barrier Loss Threshold 2.ARPV water level cannot be restored and maintained above -155 inches or cannotbe determined (Ref. H16 145 and H26 189)The reactor vessel top of active fuel was calculated in Fuel Clad Barrier Potential Loss 2.A.RCS Barrier Loss Threshold 4.ADWRRM greater than 40 R/hr.In Attachment E the detector radiation level of 40 R/hr was calculated.
Thecalculation used core inventory from NL-06-1637 to calculate isotopesconcentrations.
The calculation for DE/13I was performed to find a ratio to DEl V7Page 3 of 3Table 1 -REACTOR VESSEL LEVEL SETPOINTS AND ACTIONSSetpoints*
+54.0 (+54.5")+51.7"+42"+37"+/-32"<+30 ->+/-+20+3"-35"-60-101"-155-193"ActionTrips MAIN AND RFP TurbinesTrips HPCI and RCICHigh Level alarm from Level Recorder R608Normal Operating LevelLow Level alarm from Level Recorder R608, input to Recirculation PumpRunback to SL #2 on a loss of a RFPInput to Variable Recirculation Pump Runback to SL #4Reactor Scram, PCIS Group II, ADS Permissive (Confirmatory),
ClosesShutdown Cooling Isolation ValvesStart HPCI, RCIC and SBGT, Isolates Secondary Containment (R/F andR/B zones), PCIS Group V (RWCU), ARI Initiates Trips Reactor Recirculation PumpsPCIS Group I (MSIVs),
Start Core Spray, LPCI and Diesel Generators, ADS Permissive, Trips CRD pumps, Closes PSW isolation to TurbineBldg, Control Room Ventilation switches to Pressurization ModeTop of Active FuelContainment Spray Permissive (2/3 core coverage)
*Referenced to instrument zero.
V8, Page 1 of 3CA1 Loss of RPV Inventory.
Operability Mode Applicability:
Emergency Actuation LevelsCold Shutdown, Refueling (1 OR 2)1. Loss of RPV inventory as indicated by level less than -35" (Level 2 actuation setpoint).
According to S25213 page 12 and 14 ECCS actuates at level 2 actuation setpoint. According to H16 145 and H26189 the level 2 instruments are1/2D21N692A-D and 1/2D21N682A-D.
According to PDMS the instruments areset for -35g.2.a. RPV level cannot be monitored for 15 minutes or longerANDb. UNPLANNED level increase in any of the following due to a loss of RPVinventory.
Drywell Floor Drain SumpsDrywell Equipment drain SumpsTorusTorus room SumpsReactor Building Floor Drain SumpsTurbine Building Floor Drain SumpsRad Waste Tanks V8, Page 2 of 3CS1 Loss of RPV inventory affecting core decay heat removal capability.
Operability Mode Applicability:
Cold Shutdown, Refueling Emergency Action Levels: (1 OR 2 OR 3)1.a. Secondary CONTAINMENT INTEGRITY not established.
ANDb. RPV level less than -41" (6" below the Level 2 actuation setpoinlAccording to S25213 page 12 and 14 ECCS actuates at level 2 acthAccording to H16145 and H26 189 the level 2 instruments are 1/2D;1/2D21N682A-D.
According to PDMS the instruments are set for " -1"2.t).uation set point.21N692A-D and5". Therefore
-35"a. Secondary CONTAINMENT INTEGRITY established.
ANDb. RPV level less than -158" (TOAF).The reactor vessel top of active fuel was calculated in Fuel Clad BarrierPotential Loss 2.A. The top of fuel was found to be at vessel level of -158".3.a. RPV level cannot be monitored for 30 minutes or longer.b. Core uncovery is indicated by the following:
* UNPLANNED level increase in any of the following of sufficient magnitude to indicate core uncovery:
D Drwell Floor Drain Sumps Reactor Building Floor Drain SumpsDrywell Equipment Drain Sumps Turbine Building Floor Drain SumpsTorus Rad Waste TanksTorus Room Sumps V8Page 3 of 3Table 1 -REACTOR VESSEL LEVEL SETPOINTS AND ACTIONSSetpoints*
+54.0 (+/-54.5")+/-51.7"+/-42"+37"+32"<+30 -> +20+/-3"ActionTrips MAIN AND RFP TurbinesTrips HPCI and RCICHigh Level alarm from Level Recorder R608Normal Operating LevelLow Level alarm from Level Recorder R608, input to Recirculation PumpRunback to SL #2 on a loss of a RFPInput to Variable Recirculation Pump Runback to SL #4Reactor Scram, PCIS Group II, ADS Permissive (Confirmatory),
ClosesShutdown Cooling Isolation ValvesStart HPCI, RCIC and SBGT, Isolates Secondary Containment (R/F andR/B zones), PCIS Group V (RWCU), ARI Initiates Trips Reactor Recirculation PumpsPCIS Group I (MSIVs),
Start Core Spray, LPCI and Diesel Generators, ADS Permissive, Trips CRD pumps, Closes PSW isolation to TurbineBldg, Control Room Ventilation switches to Pressurization ModeTop of Active FuelContainment Spray Permissive (2/3 core coverage) 101"-155-193"*Referenced to instrument zero.
V9, Page 1 of 3CA1 Loss of RPV Inventory.
Operability Mode Applicability:
Cold Shutdown, Refueling Emergency Actuation Levels (1 OR 2)1. Loss of RPV inventory as indicated by level less than -35" (Level 2 actuation setpoint).
According to S25213 page 12 and 14 ECCS actuates at level 2 actuation setpoint. According to H16 145 and H26 189 the level 2 instruments are1/2D21N692A-D and 1/2D21N682A-D.
According to PDMS the instruments areset for -35".2.a. RPV level cannot be monitored for 15 minutes or longerANDb. UNPLANNED level increase in any of the following due to a loss of RPVinventory.
Drywell Floor Drain SumpsDrywell Equipment drain SumpsTorusTorus room SumpsReactor Building Floor Drain SumpsTurbine Building Floor Drain SumpsRad Waste Tanks V9, Page 2of 3CS1 Loss of RPV inventory affecting core decay heat removal capability.
Operability Mode Applicability:
Emergency Action Levels:Cold Shutdown, Refueling (1 OR20OR 3)1.a. Secondary CONTAINMENT INTEGRITY not established.
ANDb. RPV level less than -41" (6" below the Level 2 actuation setpoint).
According to S25213 page 12 and 14 ECCS actuates at level 2 actuation set point.According to H16 145 and H26 189 the level 2 instruments are 1/2D21N692A-D and1/2D21N682A-D.
According to PDMS the instruments are set for -35". Therefore
-35"-6" =-41"2.a. Secondary CONTAINMENT INTEGRITY established.
ANDb. RPV level less than -158" (TOAF).The reactor vessel top of active fuel was calculated in Fuel Clad BarrierPotential Loss 2.A. The top of fuel was found to be at vessel level of -158".3.a. RPV level cannot be monitored for 30 minutes or longer.b. Core uncovery is indicated by the following:
* UNPLANNED level increase in any of the following of sufficient magnitude to indicate core uncovery:
Drywell Floor Drain Sumps Reactor Building Floor Drain SumpsDrywell Equipment Drain Sumps Turbine Building Floor Drain SumpsTorus Rad Waste TanksTorus Room Sumps V9, Page 3of 3CG1 Loss of RPV Inventory Affecting Fuel Clad Integrity with Containment Challenged.
Operating Mode Applicability:
Emergency Action Level:Cold ShutdownRefueling (1 OR 2)1.a. RPV level less than -158" (TOAF)The reactor vessel top of active fuel was calculated in Fuel Clad BarrierPotential Loss 2.A. The top of fuel was found to be at vessel level of -158.44".
This can be rounded to -158"ANDb. ANY indication from the Containment Challenge Table C1.2.a. RPV level cannot be monitored for 30 minutes or longerANDb. Core uncovery is indicated by the following:
* UNPLANNED level increase in any of the following of sufficient magnitude to indicate core uncovery:
Drywell Floor Drain Sumps Reactor Building Floor Drain SumpsDrywell Equipment Drain Sumps Turbine Building Floor Drain SumpsTorus Rad Waste TanksTorus Room SumpsRadiation monitor readings indicative of core unco very are investigated inAttachments J and K resulting in no monitors able to provide on-scaleindications of core uncovery.
ANDc. ANY indication from the Containment Challenge Table ClContainment Challenge Table C1Containment I-t12 greater than or equal to 6% AND 02 greater than or equal to 5%Primary Containment Pressure:
greater than 56 psigSecondary CONTAINMENT INTEGRITY NOT established*
Secondary Containment radiation monitors greater than Max Safe values (SC EOP -Table 6)*If Secondary CONTAINMENT INTEGRITY is re-established prior to exceeding the 30-minute time limit, then declaration of a General Emergency is notrequired.
Damage to a loaded cask CONFINEMENT BOUNDARYE-HUI VlOPage 1 of 5Containment and Radioactivity Release Control8WROG EPGs'SAGs, Appendix BEPG/SAG Step (PCIG, continued)
Control hydrogen and oxygen concentrations in the dr~swell as follows:..........
II Ill II Illllll Illll IllI CIiI cennot be dotermined to be below 5%Sugmua Chauber Hyirogm Ce kncela< 5%Nan.Disdsd
'6%a 6%t or ceqcbedlm *,e6.t Is h&em a* o~Iet No action No sciontot be eow 6 _ _ _ _ __ _ _ _ _ _ _Control hydrogen and oxygen concentrations in the suppression chamber asfollows:Supmseo Calmber OzygenCwuontue~on a 5% or cannot be detem~ined to be below 5%OwnHl e m< 5%NaweOeidid a 86% ot owwwtSO be bedsw 6SNoracto No actionN__,____
reqired en,UE aO%orcennot
* be deterineRevision 3Revision 3 Mark l and)! containments only2-61B-16-21 VlOPage 2 of 5BWROG EPGs'SAGs, Appendix B Containment and Radioactivity Release ControlDiscussion (continued)
The third step (PC/G-3 or PC/G-6) applies when the volume is effectively deinerted (oxygen concentration equal to or greater than 5% or cannot bedetermined) and the hydrogen concentration in either volume is equal to or greaterthan 6% or cannot be determined.
Under these conditions, a potential fordeflagration exists. Venting is then permitted irrespective of the resulting radioactivity release rate and the purge may be performed using either air ornitrogen.
The recombiners are secured to eliminate a potential ignition source.A maximum of two steps, one for the drywell and one for the suppression chamber,will be performed concurrently.
Action is required,
: however, only if hydrogen isactually detected or if the concentration cannot be determined.
The specified concentrations of 6% hydrogen and 5% oxygen are the minimum valuesthat can support a deflagration.(to)
Combustion of hydrogen in the deflagration concentration range creates a traveling flame front, heating the containment atmosphere and causing a rapid increase in primary containment pressure.
Theresulting pressure peak may be high enough to rupture the primary containment ordamage the drywell-to-wetwell boundary.
Note that when oxygen concentration in one volume is equal to or greater than 5% orcannot be determined, the hydrogen concentrations in both volumes must be considered when selecting the appropriate step. If the area of concern is not inerted, hydrogen fromthe other volume could migrate to the deinerted area, creating a deflagrable mixturebefore the hydrogen monitoring system senses an increase in hydrogen concentration.
If a gas concentration cannot be determined by any means, it must be assumed to beabove the value required to support combustion.
The branch tables therefore specifysteps to perform if hydrogen or oxygen concentration cannot be determined relative toits deflagration limit. Failure or unavailability of the normal monitor,
: however, does notnecessarily mean that a gas concentration cannot be determined.
The containment isnormally inerted and hydrogen generation rates are expected to be relatively slow. If themost recent data showed considerable margin to the deflagration limits and conditions have not changed significantly since the readings were taken, it is thus unnecessary toassume that the concentrations immediately exceed the limits when direct measurement capability is lost. Rather, a decision that hydrogen and oxygen concentrations cannot bedetermined requires a judgment considering plant conditions, parameter trends, and theavailability of alternate indications.
Revision 3Revision 5 Mark Iland 11 containmenhs only 1-2B.16-23 Vl0, Page 3of 5CG1 Loss of RPV Inventory Affecting Fuel Clad Integrity with Containment Challenged.
Operating Mode Applicability:
Emergency Action Level:Cold ShutdownRefueling (1 OR 2)1.a. RPV level less than -158" (TOAF)The reactor vessel top of active fuel was calculated in Fuel Clad BarrierPotential Loss 2.A. The top of fuel was found to be at vessel level of -158.44".
This can be rounded to -158"ANDb. ANY indication from the Containment Challenge Table C1.2.a. RPV level cannot be monitored for 30 minutes or longerANDb. Core uncovery is indicated by the following:
* UNPLANNED level increase in any of the following of sufficient magnitude to indicate core uncovery:
D Drwell Floor Drain Sumps Reactor Building Floor Drain SumpsDrywell Equipment Drain Sumps Turbine Building Floor Drain SumpsTorus Rad Waste TanksTorus Room SumpsRadiation monitor readings indicative of core unco very are investigated inAttachments J and K resulting in no monitors able to provide on-scaleindications of core unco very.ANDc. ANY indication from the Containment Challenge Table Cl~Containment Challenge Table C1Containment I-2 greater than or equal to 6% AND 02 greater than or equal to 5%Primary Containment Pressure:
greater than 56 psigSecondary CONTAINMENT INTEGRITY NOT established*
Secondary Containment radiation monitors greater than Max Safe values (SC EOP -Table 6)*If Secondary CONTAINMENT INTEGRITY is re-established prior to exceeding the 30-minute time limit, then declaration of a General Emergency is notrequired.
Damage to a loaded cask CONFINEMENT BOUNDARYE-HU1 Vl0, Page 4of 52uCiigm.
GRODEC was used to calculate the fluence within the drywell.
Cylindergeometry was used to calculate the geometric fraction.
RCS Barrier Loss Threshold 5.ADrywell Fission Product Monitor reading 5.0 x 105 cpm.This EAL is to cover drywell fission product monitor indications that may indicateloss or potential loss of the RCS barrier.
Attachment F determined that thereading on drywell fission product monitor D1 1K630 of 1E6 cpm will indicatepotential loss of RCS barrier.
Per SX18062 page 34 the monitor K630 range is10 to 10^6 cpm.Primary Containment Barrier:Emergency Action Level:Primary Containment Barrier Potential Loss Threshold I .APrimary containment pressure greater than 56 psig.Containment Design Pressure:
56 psig(SS2 102005 section 304)(SS6902005 section 304)Primary Containment Barrier Potential Loss Threshold 1 .BGreater than or equal to 6% H2 AN....D 5% 02 exists inside primary containment.
Explosive mixture inside containment
> 6% Hydrogen (Ref. RG1.7pg1.7-6) z 5% Oxygen(Ref. CALC BH2-CS-52-2P33-2 pg 4 and 9)(Ref. CALC BH1-CS-33-P33-06 pg 8 & A-I)Primary Containment Barrier Potential Loss Threshold 4.ADWRRM greater than 26,000 R/hr.The evaluation of expected radiation readings on DWRRM (DI11K621) wasperformed in Attachment G of this calculation.
The detector is expected to read2. 6E4R/hr.
The range of this instrument is 1-10^7 R/hr (established in attachment C).
V10, Page 5of 527. H 16568 V5.0 "REACTOR PROTECTION SYSTEM P&ID"28. BH2-M-V999-.0047 V2.0 "DRYWELL EQUIPMENT EQ DOSES FOR EXTENDEDPOWER UPRATE FOR REA HT-96660"
: 29. HNP Technical Specifications 273/218 01-07-1630. NUREG-0016 "Calculation of releases of radioactive materials in gaseous an liquideffluents from boiling water reactors (BWR GALE Code)" April 1976.31. RG 1.183 "Alternative Radiological Source Terms for Evaluating Design BasisAccidents at Nuclear Power Reactors."
July 200032. BH2-CS-52-2P33-01 V3.0 "Containment Hydrogen Analyzer'
: 33. BH2-CS-52-2P33-02 V2.0 "Containment Oxygen Analyzer"
: 34. Regulatory Guide 1.7 "Control of Combustible Gas Concentrations in Containment Following a Loss-of-Coolant Accident" Rev 1, Sept 1976.35. SS2102005 V6.0 "Furnishing
& Delivery of Reactor Drywell & Suppression Chamber-Containment Systems"36. CALC F-86-03 V0.0 "COMPUTER CODE: VERIFICATION OF THE F GRODECCOMPUTER PROGRAM"37. Deleted38. 64CI-OCB-008-0 V8.1 "PLANT SERVICE WATER RADIATION MONITORS"
: 39. 64CI-OCB-009-0 V5.3 "LIQUID RADWASTE RADIATION MONITORING"
: 40. H 16564 V29.0 "PROCESS RADIATION MONITORING SYSTEM P&ID SHT. 2"41. H26012 V23.0 "PROCESS RADIATION MONITORING SYSTEM I.E.D. SHEET 2"42.64CI-OCB-002-1 V12.0 "UNIT ONE REACTOR BUILDING VENT RADIATION MONITORING"
: 43. 64CI-OCB-002-2 V16.0" UNIT TWO REACTOR BUILDING VENT RADIATION MONITORING"
: 44. H26013 V7.0 "PROCESS RADIATION MONITORING SYSTEM I.E.D. SHEET 3"45.64CI-OCB-003-1 V14.0 "RECOMBINER BUILDING VENT RADIATION MONITORING"
: 46. H16528 V12.0 "OFF GAS RECOMBINER BUILDING VENTILATION SYSTEM P & IDAND PROCESS FLOW DIAGRAM"47. 64CI-OCB-001-0 V13.0 "MAIN STACK RADIATION MONITORING"
: 48. S56256 V1.0 "GEI4 FUEL BUNDLE INTERFACE CONTROL"49. S54974 V0.3 "BWR SPENT FUEL STORAGE RACKS -RACK LAYOUT MPL. F16"50. S54975 V0.1 "BWR SPENT FUEL STORAGE RACKS -CONSTRUCTION (ELEVATIONS)
MPL. F16"51. H 15602 V1 .0 "REAC BLDG FUEL TRANS POUR NL PLAN SECT&DET" VllPage 1 of 3IHNP-1-FSAR-5 In the event of a process system piping failure within the drywell, reactor water and steam arereleased into the drywetl gas space. The resulting increased drywell pressure forces a mixtureof air, steam, and water through the vent system into the suppression pool. The steamcondenses rapidly in the suppression pool, resulting in rapid pressure reduction in the drywell.Air transferred during reactor blowdown to the suppression chamber pressurizes the chamberand is subsequently vented to the drywell through the vacuum relief system as the pressure inthe drywell drops below that in the suppression chamber.
Cooling systems remove heat fromthe drywell and the suppression pool for continuous cooling of the primary containment underthe postulated DBA conditions.
Isolation valves ensure the containment of radioactive materialwithin the primary containment that might be released from the reactor to the containment during course of an accident.
Other service equipment maintains the containment within Itsdesign parameters during normal operation.
The primary containment system design loadingconsiderations are provided In chapter 12 and appendix K. The safety analysis presented InHNP-2-FSAR chapter 15 demonstrates the effectiveness of the primary containment system asa radiological barrier.
In addition, primnary containment pressure and temperature transients from postulated DBAs are also discussed in HNP-2-FSAR chapter 15.6.2.2.2 DrwlThe drywell is a steel pressure vessel with a spherical lower portion 65 ft in diameter and acylindrical upper portion 35 ft 7 in. in diameter.
The overall height of the drywell Is ~111 ft. Thedesign, fabrication, inspection, and testing of the drwell comply with the requirements of theASME Code, Section III, Subsection B, Requirements for Class B Vessels, which pertains tocontainment vessels for nuclear power stations.
The primary containment is fabricated ofSA-516 grade 70 plates.IThe drywell Is designed for an internal pressure of 56 psig lcincident with a temperature of281"F, with applicable dlead, live, and seismic loads imposed on the shell. Thermal stresses inthe steel shell due to temperature gradients are also Incorporated into the design. Thus, inaccordance with the ASME Code, Section III, the maximum drywell pressure is662 psig.Although not required by the ASME Code, special precautions were taken in the fabrication ofthe steel drywell shell. Charpy V-notch specimens were used for impact testing of plate andforging material to verify proper material properties.
Plates, forgings, and pipe associated withthe drywell have an Initial nil ductility transition temperature (NDTT) < 0&deg;F when tested Inaccordance with the appropriate code for the materials.
The drywell Is assumed to be neitherpressurized nor subjected to substantial stress at temperatures below 30&deg;F.The drywell is enclosed in a reinforced concrete structure for shielding purposes.
Resistance todeformation and buckling of the drywell plate Is provided In areas where the concrete backs upthe steel shell. Above the transition zone, the drywell Is separated from the reinforced concreteby a gap of -2 In. Shielding over the top of the drywell Is provided by removable, segmented, reinforced concrete shield plugs.The removable shield plugs consist of six 3-ft-thick reinforced concrete segments spanning upto 38 ft in two separate layers of 3 segments, each weighing 180 klps. The plug segments aredesigned for 1000 Iblft uniform floor loading and were checked for the effects of the tornadoC5.2-35.2-3REV 31 9113 VllPage 2 of 3SHNP-1FSR-5 TABLE 5.2-7PRIMARY CONTAINMENT SYSTEM DESIGN PARAMETERS General Information Design PressureInternal
-dryweill~6Op1
-- -suppression chamber 58.0pi-suppression chamberDesign Temperature DrywellSuppression chamberFree VolumeDrywell (including vent system)Suppression chamber-approximate minimum-approximate maximumLeakage RateDownicomer Submergence Overall Vent Resistance Loss FactorPool Depth (Normal)No. of VentsNormal Vent Diameter (ID)Total Vent AreaNo. of Downcomers Nominal Downcomer Diameter2.0 psig2.0 psig281&deg;F146,010 ft112,900 ft115,900 ft31.2% free vol/day4 ft 0 in.(axb)4.4(0), (5.51)CaXb)12ft4 In.85ftl 11In.220 ft2802.0 ftIIa. Value is based upon Mark I Long-Tenrm Containment Program modification.
and operation in the EOD.b. Value is based upon the analysis for an RTP of 2804 M~t.c. Value is based upon original LOCA analysis.
REV 30 9/12 V11Page 3 of 3IN--FA-1,Drywell and Vent SystemsDesign internal pressure
't3QFOperating internal pressure
< 2 psig at 150&deg;F2. Suppression ChamberDesign Internal pressure 56 psig at 340&deg;FOperating intemnal pressure
< 2 psig at 50&deg; to 100&deg;FThe design internal pressure is 90% of the maximum internal pressure.
Pipe Rupture Loads (Yr, Y1, Yin)Yr= Equivalent static load on the structure generated by the reaction on thebroken high-energy pipe during the postulated break, and including anappropriate dynamic load factor to account for the dynamic nature of theload.Yj= Jet impingement equivalent static load on a structure generated by thepostulated break and including an appropriate dynamic load factor toaccount for the dynamic nature of the load.The containment is designed for the following jet impingement loadsresulting from pipe ruptures within the containment:
Area ofLocation Jet ForceInuee Drywell sphere 709,000 lb 3.94ftDuywell knuckle 472,000 lb 2.63 ft2Drywell cylinder up to el 203 ft 9 in. 472,000 lb 2.63 ft2Drywell head 32,600 lb 0.181ftThe jet forces consist of steam and/or water at 340&deg;F. Only one of theabove jet forces is considered to act in the drywell at a given time.Ym = Missile impact equivalent static load on a stru'cture generated by or duringthe postulated break, as from pipe whipping, and including an appropriate dynamic load factor to account for the dynamic nature of the load.J. Containment Flooding Loads (FL)FL = Loads generated by the post-LOCA flooding of the containment.
In theevent of a LOCA, the entire containment, including the suppression
: chamber, vent system, and the drywell, are flooded up to el 227 if, and theresulting hydrostatic load, FL, was considered In the containment design.3.8-113.8-11REV 31 9/13 V12Page 1 of 6PAGE 14 OF 30SOUTHERN NUCLEARPLANT E.I. HATCHDOCMEN TTLE IDOCUMENT NUMBER: VERSION NO:SECONDARY CONTAINMENT CONTROL 34AI3-T22-003-1 5.14ATTACHMENT 6 ATTACHMENT PAGE:TITLE: SECONDARY CONTAINMENT OPERATING RADIATION LEVELS I OF 3HVAC EXHAUST RADIATION MAXIMUM NORMAL OPERATING ANNUNCIATORS ON 1H11-P601 VALUE mr/hrHI-HI RADIATION ALARM-RX BLDG POT CONTAM AREA 1(1D11-K609A, 1D11-K609B, 1D11-K609C, 1D11-K609D) 1-REFUELING FLOOR VENT EXHAUST(1D11-K611A, 1D11-K611B, 1D11-K611C, 1D11-K611D) 1Max Normal Max SafeAREA RADIATION MONITORS Operating Operating on 1H11-P600, 1D21-P600 Value ValuemR/hr mR/hrREFUEL FLOOR AREA1 Reactor head laydown area (1 D21-K601A) 50 10002 Refueling Floor Stairway (1 D21-K601 B) 50 10003 Refueling Floor (1D21-K601D) 50 10004 Drywell Shield Plug (1D21-K601E) 50 10005 Spent Fuel Pool & New Fuel Storage (1D21-K601M) 50 1000203' ELEVATION AREA6 RB 203' Working Area (1D21-K601X) 50 1000185' ELEVATION AREA7 Spent Fuel Pool Demin. Equip (1D21-K601C) 150 10008 Fuel Pool Demain. Panel (1D21-K617) 50 100158' ELEVATION AREA9 RB 158' Working Area (1D21-K601 K) 50 100010 Rx Wtr Sample Rack Area 158' (1 D21-K601 L) 50 1000130' ELEVATION NORTH AREA11 TIP Area (1D21-K601F) 50 100012 North CRD HCU (1D21-K6O1P) 50 100013 TIP Probe Drives Area (1D21-K601 U) 100 1000MGR-0009 Rev 5.0 Vi12Page 2 of 6PAGE 15 OF 30SOUTHERN NUCLEARIPLANT E.I. HATCH _DOUMNTTILE tDOCUMENT NUMBER: VERSION NO:SECONDARY CONTAINMENT CONTROL 1 34AB-T22-003-1 5.14ATTACHMENT 6 ATTACHMENT PAGE:TITLE: SECONDARY CONTAINMENT OPERATING RADIATION LEVELS 2 OF 3Max Normal Max SafeAREA RADIATION MONITORS Operating Operating on 1 H 11 -P600, 1 D2 1-P600 Value ValuemR/hr mR/hr130' ELEVATION EAST AREA14 RB 130' N-E Working Area (1D21-K601G) 50 100015 Equipment Access Airlock (1D21-K601S) 50 1000130' ELEVATION SOUTH AREA16 RB 130' S-W Working Area (1 D21-K601 H) 50 100017 South CRD HCU (1D21-K601N) 50 1000SOUTHWEST DIAGONAL AREA18 RCIC Equip S-W Diagonal (1 D21-K601V) 50 1000NORTHWEST DIAGONAL AREA19 CRD Pump N-W Diagonal (1D21-K601W) 50 1000NORTHEAST DIAGONAL AREA20 CS & RHR N-E Diagonal (1D21-K601Y) 50 1000NORTHEAST DIAGONAL AREA21 CS & RHR S-E Diagonal (1D21-K601R) 50 1000HPCI AREA22 HPCl turbine Area (1D21-K601T) 150 1000Detector to Trip Unit cross reference DETECTOR TRIP UNITiDli-N010A, 1D11-N010B, 1D11-K609A, 1D11-K609B, 1iD11-N010C.
1D11-N010D 1 D11-K609C, 1 D11-K609D 1 D11-N012A, 1 D11-N012B, 1 D11-K6i1A, 1 D11-K611 B,1 DI1-N012C, 1 D11 -NO 12D 1 D11-K611 C, 1 D11-K611lAD 1D11-N015A
&ID11-N015B 1D11-K607A
& 1D11-K607B ID11-N016A 1D1 1-K608A1 DI1-N017A
&I1D11-N017B 1ID11-K616A
& 1D11-K616B MGR-0009 Rev 5.0 Vi 2Page 3 of 6PLANT E.I. ATCHPAGE 29 OF 30DOUMNTTILE IDOCUMENT NUMBER: VERSION NO:SECONDARY CONTAINMENT CONTROL I 34AB-T22-O03-1 5.14ATTACHMENT 10 ATTACHMENT PAGE:TITLE: SECONDARY CONTAI NMENT PARAMETERS 1 OF 1V12Page 3 of 6101 8 Pitml (13.NJ1&0))BPtmurnDIFF(1G1.D2Ct 3)103 Hx Room (1Gl1.N18) 104 HFF(,1Q3 1-N05t~N023) 05 l05 Ph FF AT DIAGONAJ AREAV~x150671806715067&#xb6;2125t6212.598212.59814" eowsr" ev.15 atxm 67' eI.~.$0U11df.~ST W)AGONAL AREA 4" akowe 3? dzow 5r7' ete',rSO)UThrEAST
[MAGONA AREA 4" diowS7elewln 14" above 87' elev.1KPcIROOU AREA 4" above r/' levion 17' above 37' .1,w,TORU$ ROOM AREA 9" above STuelev,S Table 6 CPnA1WRAL rD4 101 A PtmRnil1G31-N016C) 103 104 HK ouiD4FF 105 105 Phase Sep D01F(1e31-N2/N2E 150671506715067212.598212.598212.598on 1HtI-PUO, 1D21:.P6
____ AREA2 RueabFloorS iy(10214(501B) 50 3 Refje&~iwr9
~(1D21.K601D) 50 10004 D~wvul Shield Pbig 1D214(601E) 50 1O0005 50 1000203 ELEVATIO AREA '6 R8 203" 50 1O007 Spent Fuel PVCoo Oe Equip (1D21.K601 C) 150 10008 Fuel Pod Dem, Paule(1D21,,K617) 50 100 _9 RB1I Wdd 50 1Q0010 D21.K60L) 50 1O000 ___11 ELEVATION NOrTH ARE11 TIP ,,are(102t-KOI F) 50 1O000I2 NohCO HU (1O2l..K60lP) 50 100013 TIPv Po Dvee Area (tD2I-KB0t U) !0 10 ___014 510 N.EVATodN EArT (AREA60G) 5 114 Equipent/:&#xa2;rskudd(1D214(601S) 50 10001Eqmft30ss.EATIx*(1D2.Km 1S 50 10016 510' E. WVlorkngSU Area (01K0H 0 1017 Soth CRD HCU(1D21-KE01N) 50 1oooSOUTHWAST DIAGONAL,,REA 18 RCICEcp~
SW Diagn(1O 2t.4( )'1V) 50 1000NORTHWEST DIPGG-NAL REA- -- -19 RD Purp W DIacxaa(1D214(8OlW]
50 1000NOGll'*AST D AREA- -20 C & RIR N.qEneol (1D21.K501Y) 50 1000-O~ES DIAGONAJ NaARE21 CS&mRffS.O~eig(t1O21.W98R) 50 1O000150' ELEVATION (OT~ET22 IPCITwIoie/rea (10214011)l) 150 1000Wi-'S ''''--l' I =LM A119 Irh ShnTel ltB21.N14) 195 30&#xb6;20 01FF I6 5OPS-1933Vr4.
34AB-T22-O03-1 OPS-1 933MGR-0009 Rev 5.0 Vi12Page 4 of 6PAGE 14 OF 37SOUTHERN NUCLEARPLANT E.I. HATCHDOCMEN TTLE IDOCUMENT NUMBER: VERSION NO:SECONDARY CONTAI NM ENT CONTROL 34AB-T22-003-2 4.2ATTACHMENT 6 ATTACHMENT PAGE:TITLE: SECONDARY CONTAINMENT OPERATING RADIATION LEVELS 1 OF 4MAX NORMALHVAC EXHAUST RADIATION ANNUNCIATORS OPERATING ON 2H1 1-P601 VALUEmR/hrHI-HIREACTOR BUILDING (RX) RADIATION ANNUNCIATOR
-RX BLDG POT CONTAM AREA RADIATION (2D1 1-K609 A-D) 18REFUELING FLOOR-REFUELING FLOOR VENT EXHAUST RADIATION (2D1 1-K61 1 A-D) 18-REFUELING FLOOR VENT EXHAUST RADIATION (2D1 1-K634 A-D) 6.9-REFUELING FLOOR VENT EXHAUST RADIATION (2D1 1-K635 A-D) 5.7MAX NORMAL MAX SAFEAREA RADIATION MONITORS OPERATING OPERATING ON 2H1 1-P600, 2D21-P600 VALUE VALUEmR/hr mR/hrREFUEL FLOOR AREA1. Reactor head laydown area (2D21-K601A) 50 10002. Dryer separator pool (2D21-K601 E) 50 10003. Spent Fuel Pool & New Fuel Storage (2D21-K601M) 50 10004. Reactor Vessel Refueling Floor (2D21-K61 1K) 50 10005. Reactor Vessel Refueling Floor (2D21-K611 L) 50 1000203' ELEVATION AREA (EAST)6. CRD repair area (2D21-K601T) 50 1000203' ELEVATION AREA (WEST)7. HVAC Room West El. 203' (2D21-K600D) 50 100 Vi12Page 5 of 6PAGE 15 OF 37SOUTHERN NUCLEARPLANT E.I. HATCHDOCMEN TTLE IDOCUMENT NUMBER: VERSION NO:SECONDARY CONTAI NM ENT CONTROL 34AB-T22-003-2 4.2ATTACHMENT 6 ATTACHMENT PAGE:TITLE: SECONDARY CONTAINMENT OPERATING RADIATION LEVELS 2 OF 4MAX NORMAL MAX SAFEAREA RADIATION MONITORS OPERATING OPERATING ON 2H11-P600, 2D21-P600 VALUE VALUE____________________________________
mR/hr mR/hr185' ELEVATION AREA8. Spent fuel pool passageway (2D21-K601 P) 50 10009. RB 185' operating floor (2D21-K601R) 50 100010. RB 185' sample panel area (2D21-K601S) 50 100011. RB 185' RWCU control panel (2D21-K601U) 150 1000158' ELEVATION AREA (NORTH)12. RB 158' area N-E (2D21-K601C) 50 100013. RB 158' area N-W (2D21-K601D) 50 1000158' ELEVATION AREA (SOUTH)14. RB 158' area S-E (2D21-K601B) 50 100015. Decant pump & equipment room area 158' (2D21-K601 L) 50 1000130' ELEVATION AREA (NORTHWEST)
: 16. Tip area (2D21-K601 F) 50 1000130' ELEVATION AREA (NORTHEAST)
: 17. RB 130' N-E working area (2D21-K601G) 50 1000130' ELEVATION AREA (SOUTHEAST)
: 18. South CRD HCU (2D21-K601N) 50 1000130' ELEVATION AREA (SOUTHWEST)
: 19. RB 130' S-W working area (2D21-K601 H) 50 1000NORTHWEST DIAGONAL AREA20. RCIC equipment N-W diagonal (2D21-K601V) 50 1000SOUTHWEST DIAGONAL AREA21. CRD pump S-W diagonal (2D21-K601W) 50 1000NORTHEAST DIAGONAL AREA22. CS & RHR N-E diagonal (2D21-K601X) 50 1000SOUTHEAST DIAGONAL AREA23. CS & RHR S-E diagonal (2D21-K601Y) 150 1000 Vi12Page 6 of 6SOUTHERN NUCLEARPLANT E.I. HATCHPAGE 35 OF 37DOCMEN TILE:DOCUMENT NUMBER: VERSION NO:SECONDARY CONTAINMENT CONTROL 34AB-T22-003-2 4.2ATTACHMENT 1"1 ATTACHMENT PAGE:TITLE: SECONDARY CONTAINMENT PARAMETERS 1 OF 1I ( **t]%'I*
____T113 212,5114$ W-(2E51-N05O) 116 TorusVe 42 I e~- ~t*I~a' g CUAIVWJ A~~A 1UAIV'1 I~6 CD (021K60T)50 1000203 AREA (WT)7 HVAC RO(XWest O3'{2O21.M0GO) 50 100a Sp~tFug oo!Pa0 gway(2S.IR
: 1) 50 10(0011RB1WR' 150 100010 Swll (20214( 601D ) 50 1000)tt 8 1' 50 1000130 ELEVATION AR-A (NORTFAS) 12 R51U8 raN-E k2et 4(01C) 50 100018 R 15UhCHC 50 1Q000130 ELEVATION A (SOUTHES) 1ZODcant t4.W Zxl q ru a14(0 50 1000 _CWIONT AREA (21 Goh RD IwCU(2O214K501N) 50 1000 _ _NORTHEAST DIA GONAL AREA221CRPupS.WdagouI(22141601W) 50 1000(2D21.K601Y) 150 1000: 4A- 1I2-003-2 OPS-1 932OPS-1932 Vi13Page 1 of 5AC Sources -Operating B 3.8.1B 3.8 ELECTRICAL POWER SYSTEMSB 3.8.1 AC Sources -Operating BASESBACKGROUND The Unit 1 Class 1 E AC Electrical Power Distribution System ACsources consist of the offsite power sources (preferred powersources, normal and alternate),
and the onsite standby power sources[diesel qenerators (DGs) 1A. 1B. and lC]. As required by 10 CFR 50,Appendix A, GDC 17 (Ref. 1), the design of the AC electrical powersystem provides independence and redundancy to ensure anavailable source of power to the Engineered Safety Feature (ESF)systems.The Class I1E AC distribution system is divided into redundant loadgroups, so loss of any one group does not prevent the minimumsafety functions from being performed.
Each load group hasconnections to two preferred offsite power supplies and a single 0G.Offsite power is supplied to the 230 kV and 500 kV switchyards fromthe transmission network by eight transmission lines. From the230 kV switchyards, two electrically and physically separated circuitsprovide AC power, through startup auxiliary transformers 1C and ID,to 4.16 kV ESF buses 1E, 1F, and 1G. A detailed description of theoffsite power network and circuits to the onsite Class 1 E ESF buses isfound in the FSAR, Sections 8.3 and 8.4 (Ref. 2).An offsite circuit consists of all breakers, transformers,
: switches, interrupting
: devices, cabling, and controls required to transmit powerfrom the offsite transmission network to the onsite Class I1E ESF busor buses.Startup auxiliary transformer (SAT) 1D provides the normal source ofpower to the ESF buses 1E, 1iF, and 1G. If any 4.16 kV ESF busloses power, an automatic transfer from SAT ID to SAT IC occurs.At this time, 4.16 kV buses IA and lB and supply breakers fromSAT 1C also trip open, disconnecting all nonessential loads fromSAT 1 C to preclude overloading of the transformer.
SATs IC and ID are sized to accommodate the simultaneous startingof all required ESF loads on receipt of an accident signal without theneed for load sequencing.
: However, ESF loads are sequenced whenthe associated 4.16 kV ESF bus is supplied from SAT 1C.A description of the Unit 2 offsite power sources is provided in theBases for Unit 2 LCO 3.8.1, "AC Sources -Operating."
(continued)
HATCH UNIT I B 3.8-1 REVISION IHATCH UNIT 1B 3.8-1REVISION 1
V1 3Page 2 of 5AC Sources -Operating B 3.8.1BASESBACKGROUND fT he onsite standby power source for 4.16 kV ESF buses IE, IF, and(continued)
I1G consists of three DGs. DGs 1A and IC are dedicated to ESFI buses I1E and I1G, respectively.
DG 1 B (the swing DG) is a sharedI Power source and can supply either Unit I ESF bus 1F or Unit 2 ESF2F_ A automnatically on a of coolant accident(LOCA) signal (i.e., low reactor water level signal or high drywellpressure signal) or on an ESF bus degraded voltage or undervoltage signal. After the DG has started, it automatically ties to its respective bus after offsite power is tripped as a consequence of ESF busundervoltage or degraded
: voltage, independent of or coincident with aLOCA signal. The DGs also start and operate in the standby modewithout tying to the ESF bus on a LOCA signal alone. Following thetrip of offsite power, load shed relays strip nonpermanent loads fromthe ESF bus. When the DG is tied to the ESF bus, loads are thensequentially connected to its respective ESF bus by the automatic load sequence timing devices.
The sequencing logic controls thepermissive and starting signals to motor breakers to preventoverloading the DG.In the event of a loss of preferred power, the ESF electrical loads areautomatically connected to the DGs in sufficient time to provide forsafe reactor shutdown and to mitigate the consequences of a DesignBasis Accident (DBA) such as a LOCA.Certain required plant loads are returned to service in a predetermined sequence in order to prevent overloading of the DGs in the process.After the initiating signal is received, all automatic and permanently connected loads needed to recover the unit or maintain it in a safecondition are returned to service (i.e., the loads are energized.)
DGs IA, 1B, and IC have the following ratings:a. 2850 kW- 1000 hours, andb. 3250 kW- 168 hours.A description of the Unit 2 onsite power sources is provided in theBases for Unit 2 LCO 3.8.1.(continued)
HATCH UNIT I1 .- RVSOB 3.8-2REVISION 1
V1 3Page 3 of 5AC Sources -Operating B 3.8.1B 3.8 ELECTRICAL POWER SYSTEMSB 3.8.1 AC Sources -Operating BASESBACKGROUND The Unit 2 Class 1 E AC Electrical Power Distribution System ACsources consist of the offsite power sources (preferred powersources, normal and alternate),
and the onsite standby power sources(diesel generators (DGs) 2A, 2C, and 1 B). As required by 10 CFR 50,Appendix A, GDC 17 (Ref. 1), the design of the AC electrical powersystem provides independence and redundancy to ensure anavailable source of power to the Engineered Safety Feature (ESF)systems.The Class 1 E AC distribution system is divided into redundant loadgroups, so loss of any one group does not prevent the minimumsafety functions from being performed.
Each load group hasconnections to two preferred offsite power supplies and a single DG.Offsite power is supplied to the 230 kV and 500 kV switchyards fromthe transmission network by eight transmission lines. From the230 kV switchyards, two electrically and physically separated circuitsprovide AC power, through startup auxiliary transformers 2C and 2D,to 4.16 kV ESF buses 2E, 2F, and 2G. A detailed description of theoffsite power network and circuits to the onsite Class 1 E ESF buses isfound in the FSAR, Sections 8.2 and 8.3 (Ref. 2).An offsite circuit consists of all breakers, transformers,
: switches, interrupting
: devices, cabling, and controls required to transmit powerfrom the offsite transmission network to the onsite Class 1 E ESF busor buses.Startup auxiliary transformer (SAT) 2D provides the normal source ofpower to the ESF buses 2E, 2F, and 2G. If any 4.16 kV ESF busloses power, an automatic transfer from SAT 2D to SAT 2C occurs.At this time, 4.16 kV buses 2A and 2B and supply breakers fromSAT 2C also trip open, disconnecting all nonessential loads fromSAT 2C to preclude overloading of the transformer.
SATs 2C and 2D are sized to accommodate the simultaneous startingof all required ESF loads on receipt of an accident signal without theneed for load sequencing.
: However, ESF loads are sequenced whenthe associated 4.16 kV ESF bus is supplied from SAT 2C.A description of the Unit 1 offsite power sources is provided in theBases for Unit 1 LCO 3.8.1, "AC Sources -Operating."
(continued)
IHATCH UNIT 2 B38IRVSOB3.8-1REVISION 1
V1 3Page 4 of 5AC Sources -Operating B 3.8.1BASESBACKGROUND The onsite standby power source for 4.16 kV ESF buses 2E, 2F, and(continued) 2G consists of three DGs. DGs 2A and 2C are dedicated to ESFbuses 2E and 2G, respectively.
DB 1 B (the swing DG) is a sharedpower source and can supply either Unit I ESF bus IF or Unit 2 ESFbus 2F. A OG starts automatically on a loss of coolant accident(LOCA) signal (i.e., low reactor water level signal or high drywellpressure signal) or on an ESF bus degraded voltage or undervoltage signal. After the DG has started, it automatically ties to its respective bus after offsite power is tripped as a consequence of ESF busundervoltage or degraded
: voltage, independent of or coincident with aLOCA signal. The DGs also start and operate in the standby modewithout tying to the ESF bus on a LOCA signal alone. Following thetrip of offsite power, load shed relays strip nonpermanent loads fromthe ESF bus. When the DG is tied to the ESF bus, loads are thensequentially connected to its respective ESF bus by the automatic load sequence timing devices.
The sequencing logic controls thepermissive and starting signals to motor breakers to preventoverloading the DG.In the event of a loss of preferred power, the ESF electrical loads areautomatically connected to the DGs in sufficient time to provide forsafe reactor shutdown and to mitigate the consequences of a DesignBasis Accident (DBA) such as a LOCA.Certain required plant loads are returned to service in a predetermined sequence in order to prevent overloading of the DGs in the process.After the initiating signal is received, all automatic and permanently connected loads needed to recover the unit or maintain it in a safecondition are returned to service (i.e., the loads are energized.)
Ratings for the DGs satisfy the requirements of Regulatory Guide 1.9(Ref. 3). DGs 2A and 2C have the following ratings:a. 2850 kW -continuous,
: b. 3100 kW -2000 hours,c. 3250 kW -300 hours, andd. 3500 kW- 30 minutes.DG I1B has the following ratings:a. 2850 kW -1000 hours, and(continued)
HATCH UNIT 2 B382RVSOB 3.8-2REVISION 1
V1 3Page 5 of 52 3Okv Ma in Power Transformer No.11200 MA @ 65oC ODAFNo. 1 -A.w,,;rf105OMVA ,mo T0.88PFu(,UoTRALI
, eT,,u--' +t lUS:Sof oad o 410-Vbu es. msREV 33 9115~4160-V AUXILIARY SOUTHEN SOUITHERN NUCLEAR OPERATING COMPANY ELECTRICAL POWER SYSTEMCOMANYu~r EDINI. HATCH NUCLEAR PLANT,o.W,, o, FIGURE 8.3-1 V14Page 1 of 1Hatch -Rx Vessel Pressure Instrumentation V1 5Page 1 of 4DC Sources -Operating B 3.8.4BASESBACKGROUND (continued) result in the discharging of the associated battery (and affect thebattery cell parameters).
The DC power distribution system is described in more detail in Basesfor LCO 3.8.7, "Distribution System -Operating,"
and LCO 3.8.8,''Distribution System -Shutdown."
Each battery has adequate storage capacity to carry the required loadcontinuously for approximately 2 hours (Ref. 4).Each DC battery subsystem is separately housed in a ventilated roomapart from its charger and distribution panels. Each subsystem islocated in an area separated physically and electrically from the othersubsystems to ensure that a single failure in one subsystem does notcause a failure in a redundant subsystem.
There is no sharingbetween redundant Class 1 E subsystems such as batteries, batterychargers, or distribution panels.The batteries for DC electrical power subsystems are sized toproduce required capacity at 80% of nameplate rating, corresponding to warranted capacity at end of life. The minimum design voltage limitis 105/210 V.Each battery charger of DC electrical power subsystem has amplepower output capacity for the steady state operation of connected loads required during normal operation, while at the same timemaintaining a fully charged battery.
Each battery charger hassufficient capacity to restore the battery from the design minimumcharge to its fully charged state within 24 hours while supplying normal steady state loads (Ref. 4).A description of the Unit 2 DC power sources is provided in the Basesfor Unit 2 LCO 3.8.4, "DC Sources -Operating."
APPLICABLE SAFETY ANALYSESThe initial conditions of Design Basis Accident (DBA) and transient analyses in the FSAR, Chapters 5 and 6 (Ref. 5), and Chapter 14(Ref. 6), assume that Engineered Safety Feature (ESF) systems areOPERABLE.
The DC electrical power system provides normal andemergency DC electrical power for the DGs, emergency auxiliaries, and control and switching during all MODES of operation.
TheOPERABILITY of the DC subsystems is consistent with the initialassumptions of the accident analyses and is based upon meeting thedesign basis of the unit. This includes maintaining DC sourcesOPERABLE during accident conditions in the event of:(continued)
HATCH UNIT I1 .-3REIIN3B 3.8-53REVISION 33 V1 5Page 2 of 4HNP-1 -FSAR-88.5.3 DESCRIPTION wro separate plant batteries are furnished, each with its own static-type battery chargers, circuitIreakers, and bus. One spare battery charger is provided for each of the two batteries forenticing and to back up the two normal power supply chargers.
Plant battery operating voltageI1s/20 .lach battery with its main dc bus is in a separate room separated by a concretewall. A Class 1 ventilation system for each battery room ensures operation during emergency conditions; fire dampers are installed in the ventilation duct system to prevent the spread of firefrom one room into the other.Batteries (IA and I B) are 120-cell lead-calcium type with a continuous discharge rating of1410OAh and 1513 Ah, respectively, for 2 h at 77&deg;F to 1.75 V final average cell voltage.
Thesebatteries are not tested at the 2 hour rate.All six 125-V-dc battery chargers are full-wave silicon-controlled rectifier type rated 400 A withan output voltage regulation of + 0.75% from no load to 2% load and + 0.5% from 2% load to fullload, with ac supply variation of + 10% in voltage and + 5% in frequency.
Five separate 125-V-dc power panelboards are provided.
To maintain the required isolation and separation of the 600-V emergency
: systems, control power for each 600-V emergency busis supplied from a separate battery.
rlhe system is shown on drawing no. H-I13370.
Each of the two sets of batteries in the plant battery system has adequate storage capacity tocarry the required load for an approximate 2-h period without recharging.
A separate 125-V diesel building battery is furnished for each diesel generator and itsassociated 4-kV bus. (See drawing no. H-I13371
.) Each battery has its own SCR type batterycharger, circuit breaker, and bus with a spare battery charger for each battery to permitservicing or sparing any charger.
Emergency battery operating voltage is 125 V.Control power for each diesel generator, its generator
: breaker, and the associated 4-kVswitchgear bus power feeder circuit breakers is supplied by its respective battery.
Dieselbattery 1A also supplies control power for 4160 V switchgear bus 1 E and Division I loads on bus1 F. Diesel battery I1B also supplies emergency backup control power for 4160-V switchgear bus I F, frame 7 (RHR pump 1D). Diesel battery IC supplies control power for 4160-Vswitchgear bus 1 G and Division II loads on bus IF. Loads are as shown on figure 8.5-1.Each of the diesel building batteries has adequate storage capacity to carry the required loadfor an approximate 2-h period without recharging.
These batteries are 60-cell lead-calcium typewith a discharge rating of 410 Ah for batteries IA and 1 C and 495 Ah for battery 1 B for 8 h at77&deg;F to 1.75-V final average cell voltage.All 125-V-dc chargers are full-wave silicon-controlled rectifier type rated 100 A with a voltageregulation of + 0.75% from no load to 2% load and + 0.5% from 2% load to full load with acsupply variation of + 10% in voltage and + 5% in frequency.
8.5-28.5-2 REV 28 9/10 V1 5Page 3 of 42R23-SO03 600 V BUS C)KLm600208/1,OV125/255 VBATTERYBATrERY CHARGERSJ[)PHASE 2PRIMARY STRATEGY600 KW FLEX DGlX86-5503 ESSENTIAL 2R25CAB A5036)2R22-SO16 125 V DC BUS DIV IINSTRUMENT BUS 2A2R25-SO64 600 V MCC CPHASE 2ALTERNATE STRATEGY600 KW FLEX DG1X86-S003 cables connected directlyto charger disconnects RHR RHR MOVsRM COOLER Vi 5Page 4 of 4I BATTERY ICHARGERS IPHASE 2_______ ____PRIMARY STRATEGYESSENTIAL CAB B 125/250 V 600 KW FLEX DG2R2= .S037 BATTERY 2R22-S017 125 V DC BUS DIV II 1X86-S004 I> 600 VMCC DPHASE 2600 K'W FLEX DG))1X86-S004 cables connected directlyto charger disconnects RHR RHR MOVsINSTRUMENT BUS 2B RM COOLER2R25-S065 SMNH-1 3-021Attachment L
f..lIr~idItinn fnr F:-W~llSHEET L-6 v16Page 1 of 3CALC NO. SNCO24-CALC-007 HNP DETERMINATION OFEMERGENCY ACTION LEVEL REV. 00I E N E R C 0 N FOR INITIATING CONDITION E-PAGE NO. Page 6of 85.0 Design Inputs1. The contact dose rates from the HI-STORM 100 and HI-TRAC 125 cask systemtechnical specification
[2, Table 6.2-2] are provided below in Table 5-1. Thesesource values are scaled to develop the emergency action levels for initiating condition E-HU1.Table S-1 Technical Speification Dose Rate Limits (Neutron
+I Gamma) for HI-STORM100 and HI-TRAC 125LaatonNumber of Technical SpecIfication Measurements Limit (mrem/hr)
H I-TRAC 125 _________
Side -Mid -height 4 224.9Top 1 4 J52.8HI-STORM 100 _________
Side -60 inches below mid-height 4 38.9Side -Mid -height 4 39.7Side -60 inches above mid-height 4 15.6Top -Center of lid 1 6.0Top -Radially centered 4 8.4Inlet duct 4 72.0Outlet duct 4 18.6 Attachment LFNFRCPr4N (CAleIlItinn fnr F-HllSMNH-1 3-021SHEET L-7 v16Page 2 of 3CALC NO. SNCO24-CALC-007 HNP DETERMINATION OF -___________
EMERGENCY ACTION LEVEL REV. 0I0 E N E R C 0 N FOR INITIATING CONDITION E-" HU1 PAGE NO. Page 7oft86.0 Methodology The "on-contact"'
dose rates from the technical specification for the HI STORM-i100 andHI-TRAC 125 cask system are scaled by a factor of 2, as specified in NEI 99-01 Rev. 6[1], for use in initiating condition E-HU1.
SMNH-1 3-021Attachment LENERCON Calculation for E-HU1SHEET L-8 v16Page 3 of 3CALC NO. SNCO24-CALC-007 HNP DETERMINATION OF ___________
EMERGENCY ACTION LEVEL REV. 00O E N E R C 0 N FOR INITIATING CONDITION E- -_________HU1 PAGE NO. Page 8of 87.0 Calculations The dose rates in Table 5-1 are multiplied by 2 in order to calculate the EAL dose ratelimits. These calculations are presented below in Table 7-1.Table 7-1 Dose Rate Scaling Calculations for EAL Limits (Neutron
+ Gamma)Technical LoainSpecification Scaling Calculated Value EALLoainLimit Factor (mrem/hr)
(mrem/hr)
________________________
(mrem/hr)
____________________
_______ ______ ______ ______HI-TRAC 125 _ _ _ _ _ _ _ _ _ _ _ _ _ _Side -Mid -height 1 224.9 f 2 449.8 j 450Top J 52.8 j 2 105.6 110HI-STORM 100 ___________
Side -60 inches below mid-height 38.9 2 77.8 80Side -Mid -height 39.7 2 79.4 80Side -60 inches above mid-height 15.6 2 31.2 30Top -Center of lid 6.0 2 12 10Top -Radially centered 8.4 2 16.8 20Inlet duct 72.0 2 144 140Outlet duct 18.6 2 37.2 408.0 Computer SoftwareMicrosoft WORD 2013 is used in this calculation for basic multiplication.
V17Page 1 of 21.0 IDENTIFICATION:
ALARM PANEL 601-3______
DRY WELLHIGH PRESSUREINfITATION DEVICE:lE1 1-PIS-N694A1/BC/D 5.Aofrmta high Dyel pressure condition exists on thT4DrRwell.,4NarrowARaNge Torus tr Lv~ryweI Pres Recrdersat1pael611i 1anel2/P60.-3 52 _Enter 31ingulaP-switch RCilRPe Contropwil (NOnTWrsuli and 31omti EOSEinitiationP OPrimary Containment Control.
I-5.3 IF a high Drywell pressure condition does NOT exist and ECCS Systems have initiated, enter 34AB-E1 0-001-1, Inadvertent Initiation of ECCS/RCIC.
r-]6.0 CAUSES:S6.1 Primary system rupture inside the Drywellm6.2 Excessive N2 inerting6.3 Heatup of atmosphere In the Drywell
 
==7.0 REFERENCES==
 
18.0 TECH. sPECSJTRW/ODCM/FHA:
7.1 H1-17760 thru H1-17782, RHR System Elem j8.1 TS 3.3.5.1/ECCS Instrumentation 7.2 57CP-CAL-1 02-1, Analog Master/Slave Trip Unit Cal 8.2 TS 3.6.1 .4/Drywell Pressure34AR-601
-305-1VER. 5.2MGR-0048 Ver. 5.0 V17Page 2 of 21.0 IDENTIFICATION:
ALARM PANEL 601-3 ______DRY WELLHIGH PRESSUREINITIATION DEVICE: SETPOINT:
2E11-PIS-N694N1B/C/D 1.85 PSIG2.0 CONDITON:
13.0 CLASSIFICATION:
EMERGENCY A high pressure condition exists In the Drywell.
 
==4.0 LOCATION==
2H11 -P601 Panel 601-35.0 OPERATOR ACTIONS:5.1 Confirm a high Drywell pressure condition exists on 2T48-R607A/2T48-R607B NarrowRange Drywell Press/Torus Wtr Lvl recorders, Panel 2H1 1-P602(654).
I-5.2 IF a high Drywell pressure condition doe exist,confirm the ECCS Systems have initiated ANDDenter 31 EO-EOP-010-2, RC RPV Control (Non-ATWS).
[J5.3 IF a high Drywell pressure condition does NOT exist,enter 34AB-E1O-001-2, Inadvertent Initiation of ECCS/RCIC.
[]6.0 CAUSES:16.1 Primary system rupture inside the Drywell6.2 Excessive N2 inerting6.3 Heatup of atmosphere in the Drywell
 
==7.0 REFERENCES==
 
8.0 TECH. SPECSJTRW/ODGW/FHA:
7.1 H-27635 thru H-27657, Residual Heat Removal 8.1 TS 3.3.5.1System Eli Elementary Diagrams8.2 TS 3.5.17.2 57CP-CAL-102-2, Analog Master/Slave Trip Unit Cal34AR-601
-302-2Ver. 3.2CMGR-0048 Ver. 5.0AGMR7-10 AG-MGR-75-1101 V6Page 1 of 6SOUTHERN NUCLEAR DOCUMENT TYPE:PLANT E.I. HATCH ANNUNCIATOR RESPONSE PROCEDURE (ARP) AGE 1 OF 2DOCUMENT TITLE: DOCUMENT NUMBER:/
VERSION NO:ARP'S FOR CONTROL PANEL 1 H11-P654, 34AR-654-901-1 2.ALARM PANEL 1 2.EXPIRATION APPROVALS:
EFFECTIVE DATE: DEPARTMENT MANAGER DAK for KDL DATE 09-07-12 DATE:09-07-2012 N/A SSM /PM N/A DATE N/AANNUNCIATOR RESPONSE PROCEDURES FOR 1H11-P654
-ALARM PANEL 1ARP NO.VER NO.6.1ARP NO.VER NO.ARP NO.VER. NO.SPARE34AR-654-001
-134AR-654-031-1 4.034AR-654-061 34AR-654-002-1 1.0 34AR-654-032 SPARE 34AR-654-062-1 3.034AR-654-003-1 4.1 34AR-654-033-1 SPARE 34AR-654-063-1 2.034AR-654-004-1 6.1 34AR-654-034-1 2.0 34AR-654-064-1 3.034AR-654-005-1 3.0 34AR-654-035-1 2.0 34AR-654-065-1 7.334AR-654-006-1 7.1 34AR-654-036-1 1.0 34AR-654-066-1 2.034AR-654-007-1 12.0 34AR-654-037-1 4.0 34AR-654-067-1 8.034AR-654-008-1 4.0 34AR-654-038-1 4.0 34AR-654-068-1 8.034AR-654-009-1 5.2 34AR-654-039-1 2.0 34AR-654-069-1 2.034AR-654-0 10-1 2.1 34AR-654-040-1 2.0 34AR-654-070-1 3.134AR-654-01 1 SPARE 34AR-654-041-1 2.1 34AR-654-071-1 2.034AR-654-012-1 0.0 34AR-654-042-1 1.0 34AR-654-072-1 2.034AR-654-01 3-1 2.0 34AR-654-043-1 2.0 34AR-654-073-1 2.034AR-654-014-1 3.0 34AR-654-044-1 3.0 34AR-654-074-1 3.034AR-654-015 SPARE 34AR-654-045 SPARE 34AR-654-075-1 SPARE34AR-654-0 16-1 4.0 34AR-654-046-1 3.0 34AR-654-076-1 3.034AR-654-017 SPARE 34AR-654-047-1 1.0 34AR-654-077-1 1.034AR-654-01 8-1 3.1 34AR-654-048-1 3.0 34AR-654-078-1 1.134AR-654-01 9-1 2.0 34AR-654-049-1 3.0 34AR-654-079-1 2.034AR-654-020-1 2.0 34AR-654-050-1 4.1 34AR-654-080-1 6.234AR-654-02 1-1 5.4 34AR-654-05 1-1 3.0 34AR-654-08 1-1 SPARE34AR-654-022-1 11.1 34AR-654-052
-SPARE 34AR-654-082 SPARE34AR-654-023-1 4.0 34AR-654-053
.SPARE 34AR-654-083 SPARE34AR-654-024-1 2.0 34AR-654-054-1 3.0 34AR-654-084-1 2.034AR-654-025 SPARE 34AR-654-055-1 3.0 34AR-654-085-1 3.034AR-654-026 SPARE 34AR-654-056-1 1.1 34AR-654-086-1 2.034AR-654-027 SPARE 34AR-654-057-1 1.0 34AR-654-087-1 2.034AR-654-028-1 1.0 34AR-654-058-1 2.0 34AR-654-088-1 2.034AR-654-029-1 2.0 34AR-654-059-1 3.0 34AR-654-089-1 3.034AR-654-030 SPARE 34AR-654-060-1 0.0 34AR-654-090-1 SPARENOTE. Approval signature on this page constitutes approval for all procedures listed above at the!version indicated.
Tab numbers in the back correspond to procedure sequence number. jILevel Of Use ARPsCONTINUOUS ALLREFERENCE NoneINFONoneNMP-AP-002 V6Page 2 of 6PAGE 2 OF2SOUTHERN NUCLEARPLANT E.I. HATCHDOCUMENT TITLE: DOCUMENT NUMBER:ARP'S FOR CONTROL PANEL 1HI11-P654,I 34AR-654-901-1 ALARM PANEL 1VERSION NO:23.0UNIT 11 H11-P654
-1 (Left)654-00 1 654-002 654-003 654-004 654-005 654-006 654-007 654-008654-016 654-017 654-018 654-019 654-020 654-021 654-022 654-023654-03 1 654-032 654-033 654-034 654-035 654-036 654-037 654-038654-046 654-047 654-048 654-049 654-050 654-05 1 654-052 654-053654-06 1 654-062 654-063 654-064 654-065 654-066 654-067 654-068654-076 654-077 654-078 654-079 654-080 654-08 1 654-082 654-083UNIT I1H11-P654
-1 (Right)654-009 654-010 654-011 654-012 654-013 654-014 654-015654-024 654-025 654-026 654-027 654-028 654-029 654-030654-039 654-040 654-04 1 654-042 654-043 654-044 654-045654-054 654-055 654-056 654-057 654-058 654-059 654-060654-069 654-070 654-07 1 654-072 654-073 654-074 654-075654-084 654-085 654-086 654-087 654-088 654-089 654-090NMP-AP-002 V6Page 3 of 61.0 IDENTIFICATION:
ALARM PANEL 654SPENT FUELSTORAGE POOLLEVEL LOWDEVICE:1G41-N362 Level Sensor1G41-N372 Remote Electronics SETPOINT:
225 ft. 9 in.(21' -7" above the top of the fuel assemblies seated inthe Fuel Pool)5.0 OPERATOR ACTIONS:* Water level shall be maintained at least 21' above the top of the upper tie plates ofthe irradiated fuel assemblies seated in the fuel storage racks.Normal water level is 22' -4" to 22'- 7", as indicated on 1T24-R001.
* Water may be added to the fuel pool from the following sources:* CST, via 1G41-F041 NOTES:* Service Water, via 1G41-F217
* Fire Protection hose station* The suction piping for the DHR system has anti-siphon holes at elev. 225' 7".If level continues to drop, air could be drawn into the suction and air bind or damagethe DHR pump.5.1 Enter 34AB-G41-002-1, Decreasing Rx Well/Fuel Pool Water Level. Li5.2 IF Fuel Pool gates are installed:
5.2.1 Raise water level by regulating 1G41-F041, Spent Fuel Pool Make-up water from CSTValve, located at 185RBR07, panel 1H21-P155.
LI5.2.2 Confirm Fuel Pool Cooling filter effluent is returning to fuel pool per 34SO-G41-003-1, Fuel Pool Cooling and Cleanup System. LI5.3 IF Fuel Pool gates are NOT installed, request Maintenance to install gates. LI5.4 If the DHR system is running with suction aligned to the Unit 1 Fuel Pool,then at the direction of the SS, secure the DHR system. LI6.0 CAUSES:6.1 Water loss from normal evaporation 6.3 Malfunction of level switch (fail safe)6.2 System leakage
 
==7.0 REFERENCES==
:
8.0 TECH. SPECS./TRMIODCM/FHA 7.1 H-16002, Fuel Pool Cooling System P&ID Unit One, Section 3.7.87.2 H-17074, Fuel Pool Cooling System G41 Elem Diag34AR-654-022-IVER. 11.1MGR-0048 Ver. 5.0NMPA02 NMP-AP-002 V6Page 4 of 6DOCUMNT TTLE:DOCUMENT NUMBER: VERSION NO:ARP'SALRFOR CONTROLpAE PANEL 2H11-P654, 34AR-654-901-2 23.4EXPI RATION APPROVALS:
EFFECTIVE DATE: DEPARTMENT MANAGER DAK for KDL DATE 09-07-12 DATE:2-27-15N/A SSM /PM N/A DATE N/A _____ANNUNCIATOR RESPONSE PROCEDURES FOR 2H11-P654
-ALARM PANEL 1ARP NO. VER. NO. ARP NO. VER. NO. ARP NO. VER. NO.34AR-654-001-2 6.4 34AR-654-031-2 3.3 34AR-654-061 SPARE34AR-654-002-2 2.1 34AR-654-032-2 2.1 34AR-654-062-2 3.134AR-654-003-2 3.1 34AR-654-033 SPARE 34AR-654-063 SPARE34AR-654-004-2 2.2 34AR-654-034-2 2.3 34AR-654-064-2 2.334AR-654-005-2 2.0 34AR-654-035-2 2.1 34AR-654-065-2 4.234AR-654-006-2 8.0 34AR-654-036-2 3.4 34AR-654-066-2 5.134AR-654-007-2 11.0 34AR-654-037-2 2.0 34AR-654-067-2 6.034AR-654-008-2 3.1 34AR-654-038-2 4.1 34AR-654-068-2 6.034AR-654-009 SPARE 34AR-654-039 SPARE 34AR-654-069 SPARE34AR-654-01 0 SPARE 34AR-654-040-2 3.0 34AR-654-070-2 4.234AR-654-01 1 SPARE 34AR-654-041-2 3.1 34AR-654-071-2 4.134AR-654-012-2 3.2 34AR-654-042-2 2.1 34AR-654-072-2 4.134AR-654-013-2 2.2 34AR-654-043-2 2.1 34AR-654-073-2 4.134AR-654-014-2 1.2 34AR-654-044-2 2.1 34AR-654-074-2 5.434AR-654-015-2 3.2 34AR-654-045 SPARE 34AR-654-075 SPARE34AR-654-016-2 4.3 34AR-654-046-2 3.1 34AR-654-076-2 2.1 __34AR-654-017-2 2.1 34AR-654-047-2 3.1 34AR-654-077-2 3.1 __34AR-654-018 SPARE 34AR-654-048 SPARE 34AR-654-078-2 3.1 __34AR-654-019-2 2.2 34AR-654-049-2 1.1 34AR-654-079-2 2.234AR-654-020-2 3.2 34AR-654-050 SPARE 34AR-654-080-2 7.534AR-654-021-2 7.1 34AR-654-051 SPARE 34AR-654-081 SPARE34AR-654-022-2 11.1 34AR-654-052 SPARE 34AR-654-082 SPARE34AR-654-023-2 6.1 34AR-654-053 SPARE 34AR-654-083 SPARE34AR-654-024 SPARE 34AR-654-054 SPARE 34AR-654-084 SPARE34AR-654-025 SPARE 34AR-654-055-2 3.0 34AR-654-085-2 4.234AR-654-026-2 4.1 34AR-654-056-2 3.1 34AR-654-086-2 4.1 __34AR-654-027-2 4.0 34AR-654-057-2 3.1 34AR-654-087-2 4.1 __34AR-654-028 SPARE 34AR-654-058 SPARE 34AR-654-088-2 4.134AR-654-029 SPARE 34AR-654-059-2 2.2 34AR-654-089 SPARE34AR-654-030 SPARE 34AR-654-060 SPARE 34AR-654-090 SPAREIIIINOTQE: I Approval signature on this page constitutes approval for all procedures listed above at theIversion indicated.
Tab numbers in the back correspond to procedure sequence number.Ilit rmLevel Of Use ARPsCONTINUOUS ALLREFERENCE NoneINFO None V6Page 5 of 62H11-P654-1 (Left)654-001 654-002 654-003 654-004 654-005 654-006 654-007 654-008654-016 654-017 654-018 654-019 654-020 654-021 654-022 654-023654-031 654-032 654-033 654-034 654-035 654-036 654-037 654-038654-046 654-047 654-048 654-049 654-050 654-051 654-052 654-053654-061 654-062 654-063 654-064 654-065 654-066 654-067 654-068654-076 654-077 654-078 654-079 654-080 654-081 654-082 654-083UNIT 22H11-P654
-1 (Right)654-009 654-010 654-011 654-012 654-013 654-014 654-015654-024 654-025 654-026 654-027 654-028 654-029 654-030654-039 654-040 654-041 654-042 654-043 654-044 654-045654-054 654-055 654-056 654-057 654-058 654-059 654-060654-069 654-070 654-071 654-072 654-073 654-074 654-075654-084 654-085 654-086 654-087 654-088 654-089 654-090 V6Page 6 of 61.0 IDENTIFICATION:
ALARM PANEL 654 ________" --t---SPENT FUELSTORAGE POOLLEVEL LOWDEVICE: SETPOINT:
2G41-N372 Remote Electronics 226' 2.5" (approx.
22' 0.5" above the top of2G41-N362 Level Sensor the seated fuel assemblies)
* Water level shall be maintained at least 21' above the top of the upper tie plates of the irradiated fuel assemblies seated in the fuel storage racks.iNormal water level is 22' -4" to 22' -7" as indicated on 2T24-R001.
* Water may be added to the fuel pool from the following sources:NOTES
* CST via 2G41-F054
* Service water via 2G41-F040
* Fire Protection hose station* The suction piping for the DHR system has anti-siphon holes at elev. 225' 7".If level continues to drop, air could be drawn into the suction and air bind or damage the DHR pump.5.1 Enter 34AB-G41-002-2, Decreasing Rx Well/Fuel Pool Water Level. Li5.2 IF Fuel Pool gates are installed:
5.2.1 Raise water level by regulating 2G41-F054, Spent Fuel Pool Make-up water from CST,located at 185RBR20, panel 2H21-P155.
LI5.2.2 Confirm the Fuel Pool Cooling Filter effluent is returning to Fuel Poolper 34S0-G41-003-2, Fuel Pool Cooling and Cleanup System. Li5.2.3 Confirm air supply pressure to seals is between 33 PSIG AND 37 PSIG AND2P51-F549, 2P51-F563, and 1P51-F555 (unit one's air supply valve), Air Supply Valves,are OPEN. Li5.3 IF Fuel Pool gates are NOT INSTALLED request maintenance to INSTALL gates. Li5.4 If the DHR system is running with suction aligned to the Unit 1 Fuel Pool,then at the direction of the SS, secure the DHR system. Li6.0 CAUSES:6.1 Water loss from normal evaporation 6.5 Reactor well leakage6.2 Fuel Pool liner leakage 6.6 Dryer/Separator storage pooi leakage6.3 Fuel Pool gate leakage 6.7 Fuel Pool to transfer canal gate leakage6.4 System leakage 6.8 Malfunction of level switch (fail safe)
 
==7.0 REFERENCES==
:
18.0 TECH. SPECS./TRM/ODCM/FHA:
7.1 H-26039, Fuel Pool Cooling System P&ID Tech Specs 3.7.8, Spent Fuel Storage7.2 H-27736, Fuel Pool Cooling System Elem Pool Water Level34AR-654-022-2 VER. 11.1 V7, Page 1 of 3Fission Product Barrier Emerqency Action LevelsFuel Clad Barrier:Emergency Action LevelsFuel Clad Barrier Potential Loss Threshold 2.ARPV water level cannot be restored and maintained above -155 inches or cannotbe determined.
(Ref H16 145 and H26189)According to COLR for HNP the currently used fuel is GE 14. According to NEDC-32868P Rev 5 Appendix A (Reference of the COLR) the fuel length for GEI4 fuelwas increased from 148" to 150" inches. The Appendix A is attached below. Thusthe top of the fuel per TS Bases 2.1.1.3 is 158.44 inches below instrument zero.According to 31EO-EOP-015-1 and 31EO-EOP-015-2 "CP1 Flow Chart"operators are instructed to maximize water injection rates from alternate injection subsystems when reactor water level drops below -155 inches of instrument zero. This value is more conservative than the actual TOAF level.
V7, Page 2of 3Fuel Clad Barrier Loss Threshold 4.ADWRRM greater than 1,400 R/hr.In Attachment C the detector radiation level of 1.4E3 R/hr was calculated.
Thecalculation used core inventory from NL-06-1637 to calculate isotopesconcentrations.
The calculation for DEI131 was performed to find a ratio to DEl300uCi/gm.
GRODEC was used to calculate the fluence within the dr'ywell.
Cylinder geometry was used to calculate the geometric fraction.
Fuel Clad Barrier Loss Threshold 5.AOffgas Pre- and Post-Treatment Monitors Offscale High AND Fission ProductMonitor Offscale High.Attachment D performed an evaluation for Offgas Pre- and Post-treatment monitors D1 1-K615 (section A of Attachment D) and Containment fission productmonitors D11IP010 (section B of Attachment D). It was found that theseinstruments will be off scale.RCS Barrier:Emergency Action LevelsRCS Barrier Loss Threshold I .APrimary containment pressure greater than 1.85 psig due to RCS leakage.LIS 1C71N650A-D 1.85 psigLIS 2C71N650A-D 1.85 psigReferences (H 16568, PDMS)Ret PDMSRef. PDMSRCS Barrier Loss Threshold 2.ARPV water level cannot be restored and maintained above -155 inches or cannotbe determined (Ref. H16 145 and H26 189)The reactor vessel top of active fuel was calculated in Fuel Clad Barrier Potential Loss 2.A.RCS Barrier Loss Threshold 4.ADWRRM greater than 40 R/hr.In Attachment E the detector radiation level of 40 R/hr was calculated.
Thecalculation used core inventory from NL-06-1637 to calculate isotopesconcentrations.
The calculation for DE/13I was performed to find a ratio to DEl V7Page 3 of 3Table 1 -REACTOR VESSEL LEVEL SETPOINTS AND ACTIONSSetpoints*
+54.0 (+54.5")+51.7"+42"+37"+/-32"<+30 ->+/-+20+3"-35"-60-101"-155-193"ActionTrips MAIN AND RFP TurbinesTrips HPCI and RCICHigh Level alarm from Level Recorder R608Normal Operating LevelLow Level alarm from Level Recorder R608, input to Recirculation PumpRunback to SL #2 on a loss of a RFPInput to Variable Recirculation Pump Runback to SL #4Reactor Scram, PCIS Group II, ADS Permissive (Confirmatory),
ClosesShutdown Cooling Isolation ValvesStart HPCI, RCIC and SBGT, Isolates Secondary Containment (R/F andR/B zones), PCIS Group V (RWCU), ARI Initiates Trips Reactor Recirculation PumpsPCIS Group I (MSIVs),
Start Core Spray, LPCI and Diesel Generators, ADS Permissive, Trips CRD pumps, Closes PSW isolation to TurbineBldg, Control Room Ventilation switches to Pressurization ModeTop of Active FuelContainment Spray Permissive (2/3 core coverage)
*Referenced to instrument zero.
V8, Page 1 of 3CA1 Loss of RPV Inventory.
Operability Mode Applicability:
Emergency Actuation LevelsCold Shutdown, Refueling (1 OR 2)1. Loss of RPV inventory as indicated by level less than -35" (Level 2 actuation setpoint).
According to S25213 page 12 and 14 ECCS actuates at level 2 actuation setpoint. According to H16 145 and H26189 the level 2 instruments are1/2D21N692A-D and 1/2D21N682A-D.
According to PDMS the instruments areset for -35g.2.a. RPV level cannot be monitored for 15 minutes or longerANDb. UNPLANNED level increase in any of the following due to a loss of RPVinventory.
Drywell Floor Drain SumpsDrywell Equipment drain SumpsTorusTorus room SumpsReactor Building Floor Drain SumpsTurbine Building Floor Drain SumpsRad Waste Tanks V8, Page 2 of 3CS1 Loss of RPV inventory affecting core decay heat removal capability.
Operability Mode Applicability:
Cold Shutdown, Refueling Emergency Action Levels: (1 OR 2 OR 3)1.a. Secondary CONTAINMENT INTEGRITY not established.
ANDb. RPV level less than -41" (6" below the Level 2 actuation setpoinlAccording to S25213 page 12 and 14 ECCS actuates at level 2 acthAccording to H16145 and H26 189 the level 2 instruments are 1/2D;1/2D21N682A-D.
According to PDMS the instruments are set for " -1"2.t).uation set point.21N692A-D and5". Therefore
-35"a. Secondary CONTAINMENT INTEGRITY established.
ANDb. RPV level less than -158" (TOAF).The reactor vessel top of active fuel was calculated in Fuel Clad BarrierPotential Loss 2.A. The top of fuel was found to be at vessel level of -158".3.a. RPV level cannot be monitored for 30 minutes or longer.b. Core uncovery is indicated by the following:
* UNPLANNED level increase in any of the following of sufficient magnitude to indicate core uncovery:
D Drwell Floor Drain Sumps Reactor Building Floor Drain SumpsDrywell Equipment Drain Sumps Turbine Building Floor Drain SumpsTorus Rad Waste TanksTorus Room Sumps V8Page 3 of 3Table 1 -REACTOR VESSEL LEVEL SETPOINTS AND ACTIONSSetpoints*
+54.0 (+/-54.5")+/-51.7"+/-42"+37"+32"<+30 -> +20+/-3"ActionTrips MAIN AND RFP TurbinesTrips HPCI and RCICHigh Level alarm from Level Recorder R608Normal Operating LevelLow Level alarm from Level Recorder R608, input to Recirculation PumpRunback to SL #2 on a loss of a RFPInput to Variable Recirculation Pump Runback to SL #4Reactor Scram, PCIS Group II, ADS Permissive (Confirmatory),
ClosesShutdown Cooling Isolation ValvesStart HPCI, RCIC and SBGT, Isolates Secondary Containment (R/F andR/B zones), PCIS Group V (RWCU), ARI Initiates Trips Reactor Recirculation PumpsPCIS Group I (MSIVs),
Start Core Spray, LPCI and Diesel Generators, ADS Permissive, Trips CRD pumps, Closes PSW isolation to TurbineBldg, Control Room Ventilation switches to Pressurization ModeTop of Active FuelContainment Spray Permissive (2/3 core coverage) 101"-155-193"*Referenced to instrument zero.
V9, Page 1 of 3CA1 Loss of RPV Inventory.
Operability Mode Applicability:
Cold Shutdown, Refueling Emergency Actuation Levels (1 OR 2)1. Loss of RPV inventory as indicated by level less than -35" (Level 2 actuation setpoint).
According to S25213 page 12 and 14 ECCS actuates at level 2 actuation setpoint. According to H16 145 and H26 189 the level 2 instruments are1/2D21N692A-D and 1/2D21N682A-D.
According to PDMS the instruments areset for -35".2.a. RPV level cannot be monitored for 15 minutes or longerANDb. UNPLANNED level increase in any of the following due to a loss of RPVinventory.
Drywell Floor Drain SumpsDrywell Equipment drain SumpsTorusTorus room SumpsReactor Building Floor Drain SumpsTurbine Building Floor Drain SumpsRad Waste Tanks V9, Page 2of 3CS1 Loss of RPV inventory affecting core decay heat removal capability.
Operability Mode Applicability:
Emergency Action Levels:Cold Shutdown, Refueling (1 OR20OR 3)1.a. Secondary CONTAINMENT INTEGRITY not established.
ANDb. RPV level less than -41" (6" below the Level 2 actuation setpoint).
According to S25213 page 12 and 14 ECCS actuates at level 2 actuation set point.According to H16 145 and H26 189 the level 2 instruments are 1/2D21N692A-D and1/2D21N682A-D.
According to PDMS the instruments are set for -35". Therefore
-35"-6" =-41"2.a. Secondary CONTAINMENT INTEGRITY established.
ANDb. RPV level less than -158" (TOAF).The reactor vessel top of active fuel was calculated in Fuel Clad BarrierPotential Loss 2.A. The top of fuel was found to be at vessel level of -158".3.a. RPV level cannot be monitored for 30 minutes or longer.b. Core uncovery is indicated by the following:
* UNPLANNED level increase in any of the following of sufficient magnitude to indicate core uncovery:
Drywell Floor Drain Sumps Reactor Building Floor Drain SumpsDrywell Equipment Drain Sumps Turbine Building Floor Drain SumpsTorus Rad Waste TanksTorus Room Sumps V9, Page 3of 3CG1 Loss of RPV Inventory Affecting Fuel Clad Integrity with Containment Challenged.
Operating Mode Applicability:
Emergency Action Level:Cold ShutdownRefueling (1 OR 2)1.a. RPV level less than -158" (TOAF)The reactor vessel top of active fuel was calculated in Fuel Clad BarrierPotential Loss 2.A. The top of fuel was found to be at vessel level of -158.44".
This can be rounded to -158"ANDb. ANY indication from the Containment Challenge Table C1.2.a. RPV level cannot be monitored for 30 minutes or longerANDb. Core uncovery is indicated by the following:
* UNPLANNED level increase in any of the following of sufficient magnitude to indicate core uncovery:
Drywell Floor Drain Sumps Reactor Building Floor Drain SumpsDrywell Equipment Drain Sumps Turbine Building Floor Drain SumpsTorus Rad Waste TanksTorus Room SumpsRadiation monitor readings indicative of core unco very are investigated inAttachments J and K resulting in no monitors able to provide on-scaleindications of core uncovery.
ANDc. ANY indication from the Containment Challenge Table ClContainment Challenge Table C1Containment I-t12 greater than or equal to 6% AND 02 greater than or equal to 5%Primary Containment Pressure:
greater than 56 psigSecondary CONTAINMENT INTEGRITY NOT established*
Secondary Containment radiation monitors greater than Max Safe values (SC EOP -Table 6)*If Secondary CONTAINMENT INTEGRITY is re-established prior to exceeding the 30-minute time limit, then declaration of a General Emergency is notrequired.
Damage to a loaded cask CONFINEMENT BOUNDARYE-HUI VlOPage 1 of 5Containment and Radioactivity Release Control8WROG EPGs'SAGs, Appendix BEPG/SAG Step (PCIG, continued)
Control hydrogen and oxygen concentrations in the dr~swell as follows:..........
II Ill II Illllll Illll IllI CIiI cennot be dotermined to be below 5%Sugmua Chauber Hyirogm Ce kncela< 5%Nan.Disdsd
'6%a 6%t or ceqcbedlm *,e6.t Is h&em a* o~Iet No action No sciontot be eow 6 _ _ _ _ __ _ _ _ _ _ _Control hydrogen and oxygen concentrations in the suppression chamber asfollows:Supmseo Calmber OzygenCwuontue~on a 5% or cannot be detem~ined to be below 5%OwnHl e m< 5%NaweOeidid a 86% ot owwwtSO be bedsw 6SNoracto No actionN__,____
reqired en,UE aO%orcennot
* be deterineRevision 3Revision 3 Mark l and)! containments only2-61B-16-21 VlOPage 2 of 5BWROG EPGs'SAGs, Appendix B Containment and Radioactivity Release ControlDiscussion (continued)
The third step (PC/G-3 or PC/G-6) applies when the volume is effectively deinerted (oxygen concentration equal to or greater than 5% or cannot bedetermined) and the hydrogen concentration in either volume is equal to or greaterthan 6% or cannot be determined.
Under these conditions, a potential fordeflagration exists. Venting is then permitted irrespective of the resulting radioactivity release rate and the purge may be performed using either air ornitrogen.
The recombiners are secured to eliminate a potential ignition source.A maximum of two steps, one for the drywell and one for the suppression chamber,will be performed concurrently.
Action is required,
: however, only if hydrogen isactually detected or if the concentration cannot be determined.
The specified concentrations of 6% hydrogen and 5% oxygen are the minimum valuesthat can support a deflagration.(to)
Combustion of hydrogen in the deflagration concentration range creates a traveling flame front, heating the containment atmosphere and causing a rapid increase in primary containment pressure.
Theresulting pressure peak may be high enough to rupture the primary containment ordamage the drywell-to-wetwell boundary.
Note that when oxygen concentration in one volume is equal to or greater than 5% orcannot be determined, the hydrogen concentrations in both volumes must be considered when selecting the appropriate step. If the area of concern is not inerted, hydrogen fromthe other volume could migrate to the deinerted area, creating a deflagrable mixturebefore the hydrogen monitoring system senses an increase in hydrogen concentration.
If a gas concentration cannot be determined by any means, it must be assumed to beabove the value required to support combustion.
The branch tables therefore specifysteps to perform if hydrogen or oxygen concentration cannot be determined relative toits deflagration limit. Failure or unavailability of the normal monitor,
: however, does notnecessarily mean that a gas concentration cannot be determined.
The containment isnormally inerted and hydrogen generation rates are expected to be relatively slow. If themost recent data showed considerable margin to the deflagration limits and conditions have not changed significantly since the readings were taken, it is thus unnecessary toassume that the concentrations immediately exceed the limits when direct measurement capability is lost. Rather, a decision that hydrogen and oxygen concentrations cannot bedetermined requires a judgment considering plant conditions, parameter trends, and theavailability of alternate indications.
Revision 3Revision 5 Mark Iland 11 containmenhs only 1-2B.16-23 Vl0, Page 3of 5CG1 Loss of RPV Inventory Affecting Fuel Clad Integrity with Containment Challenged.
Operating Mode Applicability:
Emergency Action Level:Cold ShutdownRefueling (1 OR 2)1.a. RPV level less than -158" (TOAF)The reactor vessel top of active fuel was calculated in Fuel Clad BarrierPotential Loss 2.A. The top of fuel was found to be at vessel level of -158.44".
This can be rounded to -158"ANDb. ANY indication from the Containment Challenge Table C1.2.a. RPV level cannot be monitored for 30 minutes or longerANDb. Core uncovery is indicated by the following:
* UNPLANNED level increase in any of the following of sufficient magnitude to indicate core uncovery:
D Drwell Floor Drain Sumps Reactor Building Floor Drain SumpsDrywell Equipment Drain Sumps Turbine Building Floor Drain SumpsTorus Rad Waste TanksTorus Room SumpsRadiation monitor readings indicative of core unco very are investigated inAttachments J and K resulting in no monitors able to provide on-scaleindications of core unco very.ANDc. ANY indication from the Containment Challenge Table Cl~Containment Challenge Table C1Containment I-2 greater than or equal to 6% AND 02 greater than or equal to 5%Primary Containment Pressure:
greater than 56 psigSecondary CONTAINMENT INTEGRITY NOT established*
Secondary Containment radiation monitors greater than Max Safe values (SC EOP -Table 6)*If Secondary CONTAINMENT INTEGRITY is re-established prior to exceeding the 30-minute time limit, then declaration of a General Emergency is notrequired.
Damage to a loaded cask CONFINEMENT BOUNDARYE-HU1 Vl0, Page 4of 52uCiigm.
GRODEC was used to calculate the fluence within the drywell.
Cylindergeometry was used to calculate the geometric fraction.
RCS Barrier Loss Threshold 5.ADrywell Fission Product Monitor reading 5.0 x 105 cpm.This EAL is to cover drywell fission product monitor indications that may indicateloss or potential loss of the RCS barrier.
Attachment F determined that thereading on drywell fission product monitor D1 1K630 of 1E6 cpm will indicatepotential loss of RCS barrier.
Per SX18062 page 34 the monitor K630 range is10 to 10^6 cpm.Primary Containment Barrier:Emergency Action Level:Primary Containment Barrier Potential Loss Threshold I .APrimary containment pressure greater than 56 psig.Containment Design Pressure:
56 psig(SS2 102005 section 304)(SS6902005 section 304)Primary Containment Barrier Potential Loss Threshold 1 .BGreater than or equal to 6% H2 AN....D 5% 02 exists inside primary containment.
Explosive mixture inside containment
> 6% Hydrogen (Ref. RG1.7pg1.7-6) z 5% Oxygen(Ref. CALC BH2-CS-52-2P33-2 pg 4 and 9)(Ref. CALC BH1-CS-33-P33-06 pg 8 & A-I)Primary Containment Barrier Potential Loss Threshold 4.ADWRRM greater than 26,000 R/hr.The evaluation of expected radiation readings on DWRRM (DI11K621) wasperformed in Attachment G of this calculation.
The detector is expected to read2. 6E4R/hr.
The range of this instrument is 1-10^7 R/hr (established in attachment C).
V10, Page 5of 527. H 16568 V5.0 "REACTOR PROTECTION SYSTEM P&ID"28. BH2-M-V999-.0047 V2.0 "DRYWELL EQUIPMENT EQ DOSES FOR EXTENDEDPOWER UPRATE FOR REA HT-96660"
: 29. HNP Technical Specifications 273/218 01-07-1630. NUREG-0016 "Calculation of releases of radioactive materials in gaseous an liquideffluents from boiling water reactors (BWR GALE Code)" April 1976.31. RG 1.183 "Alternative Radiological Source Terms for Evaluating Design BasisAccidents at Nuclear Power Reactors."
July 200032. BH2-CS-52-2P33-01 V3.0 "Containment Hydrogen Analyzer'
: 33. BH2-CS-52-2P33-02 V2.0 "Containment Oxygen Analyzer"
: 34. Regulatory Guide 1.7 "Control of Combustible Gas Concentrations in Containment Following a Loss-of-Coolant Accident" Rev 1, Sept 1976.35. SS2102005 V6.0 "Furnishing
& Delivery of Reactor Drywell & Suppression Chamber-Containment Systems"36. CALC F-86-03 V0.0 "COMPUTER CODE: VERIFICATION OF THE F GRODECCOMPUTER PROGRAM"37. Deleted38. 64CI-OCB-008-0 V8.1 "PLANT SERVICE WATER RADIATION MONITORS"
: 39. 64CI-OCB-009-0 V5.3 "LIQUID RADWASTE RADIATION MONITORING"
: 40. H 16564 V29.0 "PROCESS RADIATION MONITORING SYSTEM P&ID SHT. 2"41. H26012 V23.0 "PROCESS RADIATION MONITORING SYSTEM I.E.D. SHEET 2"42.64CI-OCB-002-1 V12.0 "UNIT ONE REACTOR BUILDING VENT RADIATION MONITORING"
: 43. 64CI-OCB-002-2 V16.0" UNIT TWO REACTOR BUILDING VENT RADIATION MONITORING"
: 44. H26013 V7.0 "PROCESS RADIATION MONITORING SYSTEM I.E.D. SHEET 3"45.64CI-OCB-003-1 V14.0 "RECOMBINER BUILDING VENT RADIATION MONITORING"
: 46. H16528 V12.0 "OFF GAS RECOMBINER BUILDING VENTILATION SYSTEM P & IDAND PROCESS FLOW DIAGRAM"47. 64CI-OCB-001-0 V13.0 "MAIN STACK RADIATION MONITORING"
: 48. S56256 V1.0 "GEI4 FUEL BUNDLE INTERFACE CONTROL"49. S54974 V0.3 "BWR SPENT FUEL STORAGE RACKS -RACK LAYOUT MPL. F16"50. S54975 V0.1 "BWR SPENT FUEL STORAGE RACKS -CONSTRUCTION (ELEVATIONS)
MPL. F16"51. H 15602 V1 .0 "REAC BLDG FUEL TRANS POUR NL PLAN SECT&DET" VllPage 1 of 3IHNP-1-FSAR-5 In the event of a process system piping failure within the drywell, reactor water and steam arereleased into the drywetl gas space. The resulting increased drywell pressure forces a mixtureof air, steam, and water through the vent system into the suppression pool. The steamcondenses rapidly in the suppression pool, resulting in rapid pressure reduction in the drywell.Air transferred during reactor blowdown to the suppression chamber pressurizes the chamberand is subsequently vented to the drywell through the vacuum relief system as the pressure inthe drywell drops below that in the suppression chamber.
Cooling systems remove heat fromthe drywell and the suppression pool for continuous cooling of the primary containment underthe postulated DBA conditions.
Isolation valves ensure the containment of radioactive materialwithin the primary containment that might be released from the reactor to the containment during course of an accident.
Other service equipment maintains the containment within Itsdesign parameters during normal operation.
The primary containment system design loadingconsiderations are provided In chapter 12 and appendix K. The safety analysis presented InHNP-2-FSAR chapter 15 demonstrates the effectiveness of the primary containment system asa radiological barrier.
In addition, primnary containment pressure and temperature transients from postulated DBAs are also discussed in HNP-2-FSAR chapter 15.6.2.2.2 DrwlThe drywell is a steel pressure vessel with a spherical lower portion 65 ft in diameter and acylindrical upper portion 35 ft 7 in. in diameter.
The overall height of the drywell Is ~111 ft. Thedesign, fabrication, inspection, and testing of the drwell comply with the requirements of theASME Code, Section III, Subsection B, Requirements for Class B Vessels, which pertains tocontainment vessels for nuclear power stations.
The primary containment is fabricated ofSA-516 grade 70 plates.IThe drywell Is designed for an internal pressure of 56 psig lcincident with a temperature of281"F, with applicable dlead, live, and seismic loads imposed on the shell. Thermal stresses inthe steel shell due to temperature gradients are also Incorporated into the design. Thus, inaccordance with the ASME Code, Section III, the maximum drywell pressure is662 psig.Although not required by the ASME Code, special precautions were taken in the fabrication ofthe steel drywell shell. Charpy V-notch specimens were used for impact testing of plate andforging material to verify proper material properties.
Plates, forgings, and pipe associated withthe drywell have an Initial nil ductility transition temperature (NDTT) < 0&deg;F when tested Inaccordance with the appropriate code for the materials.
The drywell Is assumed to be neitherpressurized nor subjected to substantial stress at temperatures below 30&deg;F.The drywell is enclosed in a reinforced concrete structure for shielding purposes.
Resistance todeformation and buckling of the drywell plate Is provided In areas where the concrete backs upthe steel shell. Above the transition zone, the drywell Is separated from the reinforced concreteby a gap of -2 In. Shielding over the top of the drywell Is provided by removable, segmented, reinforced concrete shield plugs.The removable shield plugs consist of six 3-ft-thick reinforced concrete segments spanning upto 38 ft in two separate layers of 3 segments, each weighing 180 klps. The plug segments aredesigned for 1000 Iblft uniform floor loading and were checked for the effects of the tornadoC5.2-35.2-3REV 31 9113 VllPage 2 of 3SHNP-1FSR-5 TABLE 5.2-7PRIMARY CONTAINMENT SYSTEM DESIGN PARAMETERS General Information Design PressureInternal
-dryweill~6Op1
-- -suppression chamber 58.0pi-suppression chamberDesign Temperature DrywellSuppression chamberFree VolumeDrywell (including vent system)Suppression chamber-approximate minimum-approximate maximumLeakage RateDownicomer Submergence Overall Vent Resistance Loss FactorPool Depth (Normal)No. of VentsNormal Vent Diameter (ID)Total Vent AreaNo. of Downcomers Nominal Downcomer Diameter2.0 psig2.0 psig281&deg;F146,010 ft112,900 ft115,900 ft31.2% free vol/day4 ft 0 in.(axb)4.4(0), (5.51)CaXb)12ft4 In.85ftl 11In.220 ft2802.0 ftIIa. Value is based upon Mark I Long-Tenrm Containment Program modification.
and operation in the EOD.b. Value is based upon the analysis for an RTP of 2804 M~t.c. Value is based upon original LOCA analysis.
REV 30 9/12 V11Page 3 of 3IN--FA-1,Drywell and Vent SystemsDesign internal pressure
't3QFOperating internal pressure
< 2 psig at 150&deg;F2. Suppression ChamberDesign Internal pressure 56 psig at 340&deg;FOperating intemnal pressure
< 2 psig at 50&deg; to 100&deg;FThe design internal pressure is 90% of the maximum internal pressure.
Pipe Rupture Loads (Yr, Y1, Yin)Yr= Equivalent static load on the structure generated by the reaction on thebroken high-energy pipe during the postulated break, and including anappropriate dynamic load factor to account for the dynamic nature of theload.Yj= Jet impingement equivalent static load on a structure generated by thepostulated break and including an appropriate dynamic load factor toaccount for the dynamic nature of the load.The containment is designed for the following jet impingement loadsresulting from pipe ruptures within the containment:
Area ofLocation Jet ForceInuee Drywell sphere 709,000 lb 3.94ftDuywell knuckle 472,000 lb 2.63 ft2Drywell cylinder up to el 203 ft 9 in. 472,000 lb 2.63 ft2Drywell head 32,600 lb 0.181ftThe jet forces consist of steam and/or water at 340&deg;F. Only one of theabove jet forces is considered to act in the drywell at a given time.Ym = Missile impact equivalent static load on a stru'cture generated by or duringthe postulated break, as from pipe whipping, and including an appropriate dynamic load factor to account for the dynamic nature of the load.J. Containment Flooding Loads (FL)FL = Loads generated by the post-LOCA flooding of the containment.
In theevent of a LOCA, the entire containment, including the suppression
: chamber, vent system, and the drywell, are flooded up to el 227 if, and theresulting hydrostatic load, FL, was considered In the containment design.3.8-113.8-11REV 31 9/13 V12Page 1 of 6PAGE 14 OF 30SOUTHERN NUCLEARPLANT E.I. HATCHDOCMEN TTLE IDOCUMENT NUMBER: VERSION NO:SECONDARY CONTAINMENT CONTROL 34AI3-T22-003-1 5.14ATTACHMENT 6 ATTACHMENT PAGE:TITLE: SECONDARY CONTAINMENT OPERATING RADIATION LEVELS I OF 3HVAC EXHAUST RADIATION MAXIMUM NORMAL OPERATING ANNUNCIATORS ON 1H11-P601 VALUE mr/hrHI-HI RADIATION ALARM-RX BLDG POT CONTAM AREA 1(1D11-K609A, 1D11-K609B, 1D11-K609C, 1D11-K609D) 1-REFUELING FLOOR VENT EXHAUST(1D11-K611A, 1D11-K611B, 1D11-K611C, 1D11-K611D) 1Max Normal Max SafeAREA RADIATION MONITORS Operating Operating on 1H11-P600, 1D21-P600 Value ValuemR/hr mR/hrREFUEL FLOOR AREA1 Reactor head laydown area (1 D21-K601A) 50 10002 Refueling Floor Stairway (1 D21-K601 B) 50 10003 Refueling Floor (1D21-K601D) 50 10004 Drywell Shield Plug (1D21-K601E) 50 10005 Spent Fuel Pool & New Fuel Storage (1D21-K601M) 50 1000203' ELEVATION AREA6 RB 203' Working Area (1D21-K601X) 50 1000185' ELEVATION AREA7 Spent Fuel Pool Demin. Equip (1D21-K601C) 150 10008 Fuel Pool Demain. Panel (1D21-K617) 50 100158' ELEVATION AREA9 RB 158' Working Area (1D21-K601 K) 50 100010 Rx Wtr Sample Rack Area 158' (1 D21-K601 L) 50 1000130' ELEVATION NORTH AREA11 TIP Area (1D21-K601F) 50 100012 North CRD HCU (1D21-K6O1P) 50 100013 TIP Probe Drives Area (1D21-K601 U) 100 1000MGR-0009 Rev 5.0 Vi12Page 2 of 6PAGE 15 OF 30SOUTHERN NUCLEARIPLANT E.I. HATCH _DOUMNTTILE tDOCUMENT NUMBER: VERSION NO:SECONDARY CONTAINMENT CONTROL 1 34AB-T22-003-1 5.14ATTACHMENT 6 ATTACHMENT PAGE:TITLE: SECONDARY CONTAINMENT OPERATING RADIATION LEVELS 2 OF 3Max Normal Max SafeAREA RADIATION MONITORS Operating Operating on 1 H 11 -P600, 1 D2 1-P600 Value ValuemR/hr mR/hr130' ELEVATION EAST AREA14 RB 130' N-E Working Area (1D21-K601G) 50 100015 Equipment Access Airlock (1D21-K601S) 50 1000130' ELEVATION SOUTH AREA16 RB 130' S-W Working Area (1 D21-K601 H) 50 100017 South CRD HCU (1D21-K601N) 50 1000SOUTHWEST DIAGONAL AREA18 RCIC Equip S-W Diagonal (1 D21-K601V) 50 1000NORTHWEST DIAGONAL AREA19 CRD Pump N-W Diagonal (1D21-K601W) 50 1000NORTHEAST DIAGONAL AREA20 CS & RHR N-E Diagonal (1D21-K601Y) 50 1000NORTHEAST DIAGONAL AREA21 CS & RHR S-E Diagonal (1D21-K601R) 50 1000HPCI AREA22 HPCl turbine Area (1D21-K601T) 150 1000Detector to Trip Unit cross reference DETECTOR TRIP UNITiDli-N010A, 1D11-N010B, 1D11-K609A, 1D11-K609B, 1iD11-N010C.
1D11-N010D 1 D11-K609C, 1 D11-K609D 1 D11-N012A, 1 D11-N012B, 1 D11-K6i1A, 1 D11-K611 B,1 DI1-N012C, 1 D11 -NO 12D 1 D11-K611 C, 1 D11-K611lAD 1D11-N015A
&ID11-N015B 1D11-K607A
& 1D11-K607B ID11-N016A 1D1 1-K608A1 DI1-N017A
&I1D11-N017B 1ID11-K616A
& 1D11-K616B MGR-0009 Rev 5.0 Vi 2Page 3 of 6PLANT E.I. ATCHPAGE 29 OF 30DOUMNTTILE IDOCUMENT NUMBER: VERSION NO:SECONDARY CONTAINMENT CONTROL I 34AB-T22-O03-1 5.14ATTACHMENT 10 ATTACHMENT PAGE:TITLE: SECONDARY CONTAI NMENT PARAMETERS 1 OF 1V12Page 3 of 6101 8 Pitml (13.NJ1&0))BPtmurnDIFF(1G1.D2Ct 3)103 Hx Room (1Gl1.N18) 104 HFF(,1Q3 1-N05t~N023) 05 l05 Ph FF AT DIAGONAJ AREAV~x150671806715067&#xb6;2125t6212.598212.59814" eowsr" ev.15 atxm 67' eI.~.$0U11df.~ST W)AGONAL AREA 4" akowe 3? dzow 5r7' ete',rSO)UThrEAST
[MAGONA AREA 4" diowS7elewln 14" above 87' elev.1KPcIROOU AREA 4" above r/' levion 17' above 37' .1,w,TORU$ ROOM AREA 9" above STuelev,S Table 6 CPnA1WRAL rD4 101 A PtmRnil1G31-N016C) 103 104 HK ouiD4FF 105 105 Phase Sep D01F(1e31-N2/N2E 150671506715067212.598212.598212.598on 1HtI-PUO, 1D21:.P6
____ AREA2 RueabFloorS iy(10214(501B) 50 3 Refje&~iwr9
~(1D21.K601D) 50 10004 D~wvul Shield Pbig 1D214(601E) 50 1O0005 50 1000203 ELEVATIO AREA '6 R8 203" 50 1O007 Spent Fuel PVCoo Oe Equip (1D21.K601 C) 150 10008 Fuel Pod Dem, Paule(1D21,,K617) 50 100 _9 RB1I Wdd 50 1Q0010 D21.K60L) 50 1O000 ___11 ELEVATION NOrTH ARE11 TIP ,,are(102t-KOI F) 50 1O000I2 NohCO HU (1O2l..K60lP) 50 100013 TIPv Po Dvee Area (tD2I-KB0t U) !0 10 ___014 510 N.EVATodN EArT (AREA60G) 5 114 Equipent/:&#xa2;rskudd(1D214(601S) 50 10001Eqmft30ss.EATIx*(1D2.Km 1S 50 10016 510' E. WVlorkngSU Area (01K0H 0 1017 Soth CRD HCU(1D21-KE01N) 50 1oooSOUTHWAST DIAGONAL,,REA 18 RCICEcp~
SW Diagn(1O 2t.4( )'1V) 50 1000NORTHWEST DIPGG-NAL REA- -- -19 RD Purp W DIacxaa(1D214(8OlW]
50 1000NOGll'*AST D AREA- -20 C & RIR N.qEneol (1D21.K501Y) 50 1000-O~ES DIAGONAJ NaARE21 CS&mRffS.O~eig(t1O21.W98R) 50 1O000150' ELEVATION (OT~ET22 IPCITwIoie/rea (10214011)l) 150 1000Wi-'S ''''--l' I =LM A119 Irh ShnTel ltB21.N14) 195 30&#xb6;20 01FF I6 5OPS-1933Vr4.
34AB-T22-O03-1 OPS-1 933MGR-0009 Rev 5.0 Vi12Page 4 of 6PAGE 14 OF 37SOUTHERN NUCLEARPLANT E.I. HATCHDOCMEN TTLE IDOCUMENT NUMBER: VERSION NO:SECONDARY CONTAI NM ENT CONTROL 34AB-T22-003-2 4.2ATTACHMENT 6 ATTACHMENT PAGE:TITLE: SECONDARY CONTAINMENT OPERATING RADIATION LEVELS 1 OF 4MAX NORMALHVAC EXHAUST RADIATION ANNUNCIATORS OPERATING ON 2H1 1-P601 VALUEmR/hrHI-HIREACTOR BUILDING (RX) RADIATION ANNUNCIATOR
-RX BLDG POT CONTAM AREA RADIATION (2D1 1-K609 A-D) 18REFUELING FLOOR-REFUELING FLOOR VENT EXHAUST RADIATION (2D1 1-K61 1 A-D) 18-REFUELING FLOOR VENT EXHAUST RADIATION (2D1 1-K634 A-D) 6.9-REFUELING FLOOR VENT EXHAUST RADIATION (2D1 1-K635 A-D) 5.7MAX NORMAL MAX SAFEAREA RADIATION MONITORS OPERATING OPERATING ON 2H1 1-P600, 2D21-P600 VALUE VALUEmR/hr mR/hrREFUEL FLOOR AREA1. Reactor head laydown area (2D21-K601A) 50 10002. Dryer separator pool (2D21-K601 E) 50 10003. Spent Fuel Pool & New Fuel Storage (2D21-K601M) 50 10004. Reactor Vessel Refueling Floor (2D21-K61 1K) 50 10005. Reactor Vessel Refueling Floor (2D21-K611 L) 50 1000203' ELEVATION AREA (EAST)6. CRD repair area (2D21-K601T) 50 1000203' ELEVATION AREA (WEST)7. HVAC Room West El. 203' (2D21-K600D) 50 100 Vi12Page 5 of 6PAGE 15 OF 37SOUTHERN NUCLEARPLANT E.I. HATCHDOCMEN TTLE IDOCUMENT NUMBER: VERSION NO:SECONDARY CONTAI NM ENT CONTROL 34AB-T22-003-2 4.2ATTACHMENT 6 ATTACHMENT PAGE:TITLE: SECONDARY CONTAINMENT OPERATING RADIATION LEVELS 2 OF 4MAX NORMAL MAX SAFEAREA RADIATION MONITORS OPERATING OPERATING ON 2H11-P600, 2D21-P600 VALUE VALUE____________________________________
mR/hr mR/hr185' ELEVATION AREA8. Spent fuel pool passageway (2D21-K601 P) 50 10009. RB 185' operating floor (2D21-K601R) 50 100010. RB 185' sample panel area (2D21-K601S) 50 100011. RB 185' RWCU control panel (2D21-K601U) 150 1000158' ELEVATION AREA (NORTH)12. RB 158' area N-E (2D21-K601C) 50 100013. RB 158' area N-W (2D21-K601D) 50 1000158' ELEVATION AREA (SOUTH)14. RB 158' area S-E (2D21-K601B) 50 100015. Decant pump & equipment room area 158' (2D21-K601 L) 50 1000130' ELEVATION AREA (NORTHWEST)
: 16. Tip area (2D21-K601 F) 50 1000130' ELEVATION AREA (NORTHEAST)
: 17. RB 130' N-E working area (2D21-K601G) 50 1000130' ELEVATION AREA (SOUTHEAST)
: 18. South CRD HCU (2D21-K601N) 50 1000130' ELEVATION AREA (SOUTHWEST)
: 19. RB 130' S-W working area (2D21-K601 H) 50 1000NORTHWEST DIAGONAL AREA20. RCIC equipment N-W diagonal (2D21-K601V) 50 1000SOUTHWEST DIAGONAL AREA21. CRD pump S-W diagonal (2D21-K601W) 50 1000NORTHEAST DIAGONAL AREA22. CS & RHR N-E diagonal (2D21-K601X) 50 1000SOUTHEAST DIAGONAL AREA23. CS & RHR S-E diagonal (2D21-K601Y) 150 1000 Vi12Page 6 of 6SOUTHERN NUCLEARPLANT E.I. HATCHPAGE 35 OF 37DOCMEN TILE:DOCUMENT NUMBER: VERSION NO:SECONDARY CONTAINMENT CONTROL 34AB-T22-003-2 4.2ATTACHMENT 1"1 ATTACHMENT PAGE:TITLE: SECONDARY CONTAINMENT PARAMETERS 1 OF 1I ( **t]%'I*
____T113 212,5114$ W-(2E51-N05O) 116 TorusVe 42 I e~- ~t*I~a' g CUAIVWJ A~~A 1UAIV'1 I~6 CD (021K60T)50 1000203 AREA (WT)7 HVAC RO(XWest O3'{2O21.M0GO) 50 100a Sp~tFug oo!Pa0 gway(2S.IR
: 1) 50 10(0011RB1WR' 150 100010 Swll (20214( 601D ) 50 1000)tt 8 1' 50 1000130 ELEVATION AR-A (NORTFAS) 12 R51U8 raN-E k2et 4(01C) 50 100018 R 15UhCHC 50 1Q000130 ELEVATION A (SOUTHES) 1ZODcant t4.W Zxl q ru a14(0 50 1000 _CWIONT AREA (21 Goh RD IwCU(2O214K501N) 50 1000 _ _NORTHEAST DIA GONAL AREA221CRPupS.WdagouI(22141601W) 50 1000(2D21.K601Y) 150 1000: 4A- 1I2-003-2 OPS-1 932OPS-1932 Vi13Page 1 of 5AC Sources -Operating B 3.8.1B 3.8 ELECTRICAL POWER SYSTEMSB 3.8.1 AC Sources -Operating BASESBACKGROUND The Unit 1 Class 1 E AC Electrical Power Distribution System ACsources consist of the offsite power sources (preferred powersources, normal and alternate),
and the onsite standby power sources[diesel qenerators (DGs) 1A. 1B. and lC]. As required by 10 CFR 50,Appendix A, GDC 17 (Ref. 1), the design of the AC electrical powersystem provides independence and redundancy to ensure anavailable source of power to the Engineered Safety Feature (ESF)systems.The Class I1E AC distribution system is divided into redundant loadgroups, so loss of any one group does not prevent the minimumsafety functions from being performed.
Each load group hasconnections to two preferred offsite power supplies and a single 0G.Offsite power is supplied to the 230 kV and 500 kV switchyards fromthe transmission network by eight transmission lines. From the230 kV switchyards, two electrically and physically separated circuitsprovide AC power, through startup auxiliary transformers 1C and ID,to 4.16 kV ESF buses 1E, 1F, and 1G. A detailed description of theoffsite power network and circuits to the onsite Class 1 E ESF buses isfound in the FSAR, Sections 8.3 and 8.4 (Ref. 2).An offsite circuit consists of all breakers, transformers,
: switches, interrupting
: devices, cabling, and controls required to transmit powerfrom the offsite transmission network to the onsite Class I1E ESF busor buses.Startup auxiliary transformer (SAT) 1D provides the normal source ofpower to the ESF buses 1E, 1iF, and 1G. If any 4.16 kV ESF busloses power, an automatic transfer from SAT ID to SAT IC occurs.At this time, 4.16 kV buses IA and lB and supply breakers fromSAT 1C also trip open, disconnecting all nonessential loads fromSAT 1 C to preclude overloading of the transformer.
SATs IC and ID are sized to accommodate the simultaneous startingof all required ESF loads on receipt of an accident signal without theneed for load sequencing.
: However, ESF loads are sequenced whenthe associated 4.16 kV ESF bus is supplied from SAT 1C.A description of the Unit 2 offsite power sources is provided in theBases for Unit 2 LCO 3.8.1, "AC Sources -Operating."
(continued)
HATCH UNIT I B 3.8-1 REVISION IHATCH UNIT 1B 3.8-1REVISION 1
V1 3Page 2 of 5AC Sources -Operating B 3.8.1BASESBACKGROUND fT he onsite standby power source for 4.16 kV ESF buses IE, IF, and(continued)
I1G consists of three DGs. DGs 1A and IC are dedicated to ESFI buses I1E and I1G, respectively.
DG 1 B (the swing DG) is a sharedI Power source and can supply either Unit I ESF bus 1F or Unit 2 ESF2F_ A automnatically on a of coolant accident(LOCA) signal (i.e., low reactor water level signal or high drywellpressure signal) or on an ESF bus degraded voltage or undervoltage signal. After the DG has started, it automatically ties to its respective bus after offsite power is tripped as a consequence of ESF busundervoltage or degraded
: voltage, independent of or coincident with aLOCA signal. The DGs also start and operate in the standby modewithout tying to the ESF bus on a LOCA signal alone. Following thetrip of offsite power, load shed relays strip nonpermanent loads fromthe ESF bus. When the DG is tied to the ESF bus, loads are thensequentially connected to its respective ESF bus by the automatic load sequence timing devices.
The sequencing logic controls thepermissive and starting signals to motor breakers to preventoverloading the DG.In the event of a loss of preferred power, the ESF electrical loads areautomatically connected to the DGs in sufficient time to provide forsafe reactor shutdown and to mitigate the consequences of a DesignBasis Accident (DBA) such as a LOCA.Certain required plant loads are returned to service in a predetermined sequence in order to prevent overloading of the DGs in the process.After the initiating signal is received, all automatic and permanently connected loads needed to recover the unit or maintain it in a safecondition are returned to service (i.e., the loads are energized.)
DGs IA, 1B, and IC have the following ratings:a. 2850 kW- 1000 hours, andb. 3250 kW- 168 hours.A description of the Unit 2 onsite power sources is provided in theBases for Unit 2 LCO 3.8.1.(continued)
HATCH UNIT I1 .- RVSOB 3.8-2REVISION 1
V1 3Page 3 of 5AC Sources -Operating B 3.8.1B 3.8 ELECTRICAL POWER SYSTEMSB 3.8.1 AC Sources -Operating BASESBACKGROUND The Unit 2 Class 1 E AC Electrical Power Distribution System ACsources consist of the offsite power sources (preferred powersources, normal and alternate),
and the onsite standby power sources(diesel generators (DGs) 2A, 2C, and 1 B). As required by 10 CFR 50,Appendix A, GDC 17 (Ref. 1), the design of the AC electrical powersystem provides independence and redundancy to ensure anavailable source of power to the Engineered Safety Feature (ESF)systems.The Class 1 E AC distribution system is divided into redundant loadgroups, so loss of any one group does not prevent the minimumsafety functions from being performed.
Each load group hasconnections to two preferred offsite power supplies and a single DG.Offsite power is supplied to the 230 kV and 500 kV switchyards fromthe transmission network by eight transmission lines. From the230 kV switchyards, two electrically and physically separated circuitsprovide AC power, through startup auxiliary transformers 2C and 2D,to 4.16 kV ESF buses 2E, 2F, and 2G. A detailed description of theoffsite power network and circuits to the onsite Class 1 E ESF buses isfound in the FSAR, Sections 8.2 and 8.3 (Ref. 2).An offsite circuit consists of all breakers, transformers,
: switches, interrupting
: devices, cabling, and controls required to transmit powerfrom the offsite transmission network to the onsite Class 1 E ESF busor buses.Startup auxiliary transformer (SAT) 2D provides the normal source ofpower to the ESF buses 2E, 2F, and 2G. If any 4.16 kV ESF busloses power, an automatic transfer from SAT 2D to SAT 2C occurs.At this time, 4.16 kV buses 2A and 2B and supply breakers fromSAT 2C also trip open, disconnecting all nonessential loads fromSAT 2C to preclude overloading of the transformer.
SATs 2C and 2D are sized to accommodate the simultaneous startingof all required ESF loads on receipt of an accident signal without theneed for load sequencing.
: However, ESF loads are sequenced whenthe associated 4.16 kV ESF bus is supplied from SAT 2C.A description of the Unit 1 offsite power sources is provided in theBases for Unit 1 LCO 3.8.1, "AC Sources -Operating."
(continued)
IHATCH UNIT 2 B38IRVSOB3.8-1REVISION 1
V1 3Page 4 of 5AC Sources -Operating B 3.8.1BASESBACKGROUND The onsite standby power source for 4.16 kV ESF buses 2E, 2F, and(continued) 2G consists of three DGs. DGs 2A and 2C are dedicated to ESFbuses 2E and 2G, respectively.
DB 1 B (the swing DG) is a sharedpower source and can supply either Unit I ESF bus IF or Unit 2 ESFbus 2F. A OG starts automatically on a loss of coolant accident(LOCA) signal (i.e., low reactor water level signal or high drywellpressure signal) or on an ESF bus degraded voltage or undervoltage signal. After the DG has started, it automatically ties to its respective bus after offsite power is tripped as a consequence of ESF busundervoltage or degraded
: voltage, independent of or coincident with aLOCA signal. The DGs also start and operate in the standby modewithout tying to the ESF bus on a LOCA signal alone. Following thetrip of offsite power, load shed relays strip nonpermanent loads fromthe ESF bus. When the DG is tied to the ESF bus, loads are thensequentially connected to its respective ESF bus by the automatic load sequence timing devices.
The sequencing logic controls thepermissive and starting signals to motor breakers to preventoverloading the DG.In the event of a loss of preferred power, the ESF electrical loads areautomatically connected to the DGs in sufficient time to provide forsafe reactor shutdown and to mitigate the consequences of a DesignBasis Accident (DBA) such as a LOCA.Certain required plant loads are returned to service in a predetermined sequence in order to prevent overloading of the DGs in the process.After the initiating signal is received, all automatic and permanently connected loads needed to recover the unit or maintain it in a safecondition are returned to service (i.e., the loads are energized.)
Ratings for the DGs satisfy the requirements of Regulatory Guide 1.9(Ref. 3). DGs 2A and 2C have the following ratings:a. 2850 kW -continuous,
: b. 3100 kW -2000 hours,c. 3250 kW -300 hours, andd. 3500 kW- 30 minutes.DG I1B has the following ratings:a. 2850 kW -1000 hours, and(continued)
HATCH UNIT 2 B382RVSOB 3.8-2REVISION 1
V1 3Page 5 of 52 3Okv Ma in Power Transformer No.11200 MA @ 65oC ODAFNo. 1 -A.w,,;rf105OMVA ,mo T0.88PFu(,UoTRALI
, eT,,u--' +t lUS:Sof oad o 410-Vbu es. msREV 33 9115~4160-V AUXILIARY SOUTHEN SOUITHERN NUCLEAR OPERATING COMPANY ELECTRICAL POWER SYSTEMCOMANYu~r EDINI. HATCH NUCLEAR PLANT,o.W,, o, FIGURE 8.3-1 V14Page 1 of 1Hatch -Rx Vessel Pressure Instrumentation V1 5Page 1 of 4DC Sources -Operating B 3.8.4BASESBACKGROUND (continued) result in the discharging of the associated battery (and affect thebattery cell parameters).
The DC power distribution system is described in more detail in Basesfor LCO 3.8.7, "Distribution System -Operating,"
and LCO 3.8.8,''Distribution System -Shutdown."
Each battery has adequate storage capacity to carry the required loadcontinuously for approximately 2 hours (Ref. 4).Each DC battery subsystem is separately housed in a ventilated roomapart from its charger and distribution panels. Each subsystem islocated in an area separated physically and electrically from the othersubsystems to ensure that a single failure in one subsystem does notcause a failure in a redundant subsystem.
There is no sharingbetween redundant Class 1 E subsystems such as batteries, batterychargers, or distribution panels.The batteries for DC electrical power subsystems are sized toproduce required capacity at 80% of nameplate rating, corresponding to warranted capacity at end of life. The minimum design voltage limitis 105/210 V.Each battery charger of DC electrical power subsystem has amplepower output capacity for the steady state operation of connected loads required during normal operation, while at the same timemaintaining a fully charged battery.
Each battery charger hassufficient capacity to restore the battery from the design minimumcharge to its fully charged state within 24 hours while supplying normal steady state loads (Ref. 4).A description of the Unit 2 DC power sources is provided in the Basesfor Unit 2 LCO 3.8.4, "DC Sources -Operating."
APPLICABLE SAFETY ANALYSESThe initial conditions of Design Basis Accident (DBA) and transient analyses in the FSAR, Chapters 5 and 6 (Ref. 5), and Chapter 14(Ref. 6), assume that Engineered Safety Feature (ESF) systems areOPERABLE.
The DC electrical power system provides normal andemergency DC electrical power for the DGs, emergency auxiliaries, and control and switching during all MODES of operation.
TheOPERABILITY of the DC subsystems is consistent with the initialassumptions of the accident analyses and is based upon meeting thedesign basis of the unit. This includes maintaining DC sourcesOPERABLE during accident conditions in the event of:(continued)
HATCH UNIT I1 .-3REIIN3B 3.8-53REVISION 33 V1 5Page 2 of 4HNP-1 -FSAR-88.5.3 DESCRIPTION wro separate plant batteries are furnished, each with its own static-type battery chargers, circuitIreakers, and bus. One spare battery charger is provided for each of the two batteries forenticing and to back up the two normal power supply chargers.
Plant battery operating voltageI1s/20 .lach battery with its main dc bus is in a separate room separated by a concretewall. A Class 1 ventilation system for each battery room ensures operation during emergency conditions; fire dampers are installed in the ventilation duct system to prevent the spread of firefrom one room into the other.Batteries (IA and I B) are 120-cell lead-calcium type with a continuous discharge rating of1410OAh and 1513 Ah, respectively, for 2 h at 77&deg;F to 1.75 V final average cell voltage.
Thesebatteries are not tested at the 2 hour rate.All six 125-V-dc battery chargers are full-wave silicon-controlled rectifier type rated 400 A withan output voltage regulation of + 0.75% from no load to 2% load and + 0.5% from 2% load to fullload, with ac supply variation of + 10% in voltage and + 5% in frequency.
Five separate 125-V-dc power panelboards are provided.
To maintain the required isolation and separation of the 600-V emergency
: systems, control power for each 600-V emergency busis supplied from a separate battery.
rlhe system is shown on drawing no. H-I13370.
Each of the two sets of batteries in the plant battery system has adequate storage capacity tocarry the required load for an approximate 2-h period without recharging.
A separate 125-V diesel building battery is furnished for each diesel generator and itsassociated 4-kV bus. (See drawing no. H-I13371
.) Each battery has its own SCR type batterycharger, circuit breaker, and bus with a spare battery charger for each battery to permitservicing or sparing any charger.
Emergency battery operating voltage is 125 V.Control power for each diesel generator, its generator
: breaker, and the associated 4-kVswitchgear bus power feeder circuit breakers is supplied by its respective battery.
Dieselbattery 1A also supplies control power for 4160 V switchgear bus 1 E and Division I loads on bus1 F. Diesel battery I1B also supplies emergency backup control power for 4160-V switchgear bus I F, frame 7 (RHR pump 1D). Diesel battery IC supplies control power for 4160-Vswitchgear bus 1 G and Division II loads on bus IF. Loads are as shown on figure 8.5-1.Each of the diesel building batteries has adequate storage capacity to carry the required loadfor an approximate 2-h period without recharging.
These batteries are 60-cell lead-calcium typewith a discharge rating of 410 Ah for batteries IA and 1 C and 495 Ah for battery 1 B for 8 h at77&deg;F to 1.75-V final average cell voltage.All 125-V-dc chargers are full-wave silicon-controlled rectifier type rated 100 A with a voltageregulation of + 0.75% from no load to 2% load and + 0.5% from 2% load to full load with acsupply variation of + 10% in voltage and + 5% in frequency.
8.5-28.5-2 REV 28 9/10 V1 5Page 3 of 42R23-SO03 600 V BUS C)KLm600208/1,OV125/255 VBATTERYBATrERY CHARGERSJ[)PHASE 2PRIMARY STRATEGY600 KW FLEX DGlX86-5503 ESSENTIAL 2R25CAB A5036)2R22-SO16 125 V DC BUS DIV IINSTRUMENT BUS 2A2R25-SO64 600 V MCC CPHASE 2ALTERNATE STRATEGY600 KW FLEX DG1X86-S003 cables connected directlyto charger disconnects RHR RHR MOVsRM COOLER Vi 5Page 4 of 4I BATTERY ICHARGERS IPHASE 2_______ ____PRIMARY STRATEGYESSENTIAL CAB B 125/250 V 600 KW FLEX DG2R2= .S037 BATTERY 2R22-S017 125 V DC BUS DIV II 1X86-S004 I> 600 VMCC DPHASE 2600 K'W FLEX DG))1X86-S004 cables connected directlyto charger disconnects RHR RHR MOVsINSTRUMENT BUS 2B RM COOLER2R25-S065 SMNH-1 3-021Attachment L
f..lIr~idItinn fnr F:-W~llSHEET L-6 v16Page 1 of 3CALC NO. SNCO24-CALC-007 HNP DETERMINATION OFEMERGENCY ACTION LEVEL REV. 00I E N E R C 0 N FOR INITIATING CONDITION E-PAGE NO. Page 6of 85.0 Design Inputs1. The contact dose rates from the HI-STORM 100 and HI-TRAC 125 cask systemtechnical specification
[2, Table 6.2-2] are provided below in Table 5-1. Thesesource values are scaled to develop the emergency action levels for initiating condition E-HU1.Table S-1 Technical Speification Dose Rate Limits (Neutron
+I Gamma) for HI-STORM100 and HI-TRAC 125LaatonNumber of Technical SpecIfication Measurements Limit (mrem/hr)
H I-TRAC 125 _________
Side -Mid -height 4 224.9Top 1 4 J52.8HI-STORM 100 _________
Side -60 inches below mid-height 4 38.9Side -Mid -height 4 39.7Side -60 inches above mid-height 4 15.6Top -Center of lid 1 6.0Top -Radially centered 4 8.4Inlet duct 4 72.0Outlet duct 4 18.6 Attachment LFNFRCPr4N (CAleIlItinn fnr F-HllSMNH-1 3-021SHEET L-7 v16Page 2 of 3CALC NO. SNCO24-CALC-007 HNP DETERMINATION OF -___________
EMERGENCY ACTION LEVEL REV. 0I0 E N E R C 0 N FOR INITIATING CONDITION E-" HU1 PAGE NO. Page 7oft86.0 Methodology The "on-contact"'
dose rates from the technical specification for the HI STORM-i100 andHI-TRAC 125 cask system are scaled by a factor of 2, as specified in NEI 99-01 Rev. 6[1], for use in initiating condition E-HU1.
SMNH-1 3-021Attachment LENERCON Calculation for E-HU1SHEET L-8 v16Page 3 of 3CALC NO. SNCO24-CALC-007 HNP DETERMINATION OF ___________
EMERGENCY ACTION LEVEL REV. 00O E N E R C 0 N FOR INITIATING CONDITION E- -_________HU1 PAGE NO. Page 8of 87.0 Calculations The dose rates in Table 5-1 are multiplied by 2 in order to calculate the EAL dose ratelimits. These calculations are presented below in Table 7-1.Table 7-1 Dose Rate Scaling Calculations for EAL Limits (Neutron
+ Gamma)Technical LoainSpecification Scaling Calculated Value EALLoainLimit Factor (mrem/hr)
(mrem/hr)
________________________
(mrem/hr)
____________________
_______ ______ ______ ______HI-TRAC 125 _ _ _ _ _ _ _ _ _ _ _ _ _ _Side -Mid -height 1 224.9 f 2 449.8 j 450Top J 52.8 j 2 105.6 110HI-STORM 100 ___________
Side -60 inches below mid-height 38.9 2 77.8 80Side -Mid -height 39.7 2 79.4 80Side -60 inches above mid-height 15.6 2 31.2 30Top -Center of lid 6.0 2 12 10Top -Radially centered 8.4 2 16.8 20Inlet duct 72.0 2 144 140Outlet duct 18.6 2 37.2 408.0 Computer SoftwareMicrosoft WORD 2013 is used in this calculation for basic multiplication.
V17Page 1 of 21.0 IDENTIFICATION:
ALARM PANEL 601-3______
DRY WELLHIGH PRESSUREINfITATION DEVICE:lE1 1-PIS-N694A1/BC/D 5.Aofrmta high Dyel pressure condition exists on thT4DrRwell.,4NarrowARaNge Torus tr Lv~ryweI Pres Recrdersat1pael611i 1anel2/P60.-3 52 _Enter 31ingulaP-switch RCilRPe Contropwil (NOnTWrsuli and 31omti EOSEinitiationP OPrimary Containment Control.
I-5.3 IF a high Drywell pressure condition does NOT exist and ECCS Systems have initiated, enter 34AB-E1 0-001-1, Inadvertent Initiation of ECCS/RCIC.
r-]6.0 CAUSES:S6.1 Primary system rupture inside the Drywellm6.2 Excessive N2 inerting6.3 Heatup of atmosphere In the Drywell
 
==7.0 REFERENCES==
 
18.0 TECH. sPECSJTRW/ODCM/FHA:
7.1 H1-17760 thru H1-17782, RHR System Elem j8.1 TS 3.3.5.1/ECCS Instrumentation 7.2 57CP-CAL-1 02-1, Analog Master/Slave Trip Unit Cal 8.2 TS 3.6.1 .4/Drywell Pressure34AR-601
-305-1VER. 5.2MGR-0048 Ver. 5.0 V17Page 2 of 21.0 IDENTIFICATION:
ALARM PANEL 601-3 ______DRY WELLHIGH PRESSUREINITIATION DEVICE: SETPOINT:
2E11-PIS-N694N1B/C/D 1.85 PSIG2.0 CONDITON:
13.0 CLASSIFICATION:
EMERGENCY A high pressure condition exists In the Drywell.
 
==4.0 LOCATION==
2H11 -P601 Panel 601-35.0 OPERATOR ACTIONS:5.1 Confirm a high Drywell pressure condition exists on 2T48-R607A/2T48-R607B NarrowRange Drywell Press/Torus Wtr Lvl recorders, Panel 2H1 1-P602(654).
I-5.2 IF a high Drywell pressure condition doe exist,confirm the ECCS Systems have initiated ANDDenter 31 EO-EOP-010-2, RC RPV Control (Non-ATWS).
[J5.3 IF a high Drywell pressure condition does NOT exist,enter 34AB-E1O-001-2, Inadvertent Initiation of ECCS/RCIC.
[]6.0 CAUSES:16.1 Primary system rupture inside the Drywell6.2 Excessive N2 inerting6.3 Heatup of atmosphere in the Drywell
 
==7.0 REFERENCES==
 
8.0 TECH. SPECSJTRW/ODGW/FHA:
7.1 H-27635 thru H-27657, Residual Heat Removal 8.1 TS 3.3.5.1System Eli Elementary Diagrams8.2 TS 3.5.17.2 57CP-CAL-102-2, Analog Master/Slave Trip Unit Cal34AR-601
-302-2Ver. 3.2CMGR-0048 Ver. 5.0AGMR7-10 AG-MGR-75-1101}}

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