Text

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
	ML16071A150
Person / Time
Site:	Hatch, Vogtle, Farley
Issue date:	03/03/2016
From:	; Southern Nuclear Operating Co
To:	; Office of Nuclear Reactor Regulation
Shared Package
ML16071A108	List: ML16071A110; ML16071A126; ML16071A132; ML16071A136; ML16071A137; ML16071A140; ML16071A144; ML16071A146; ML16071A147; ML16071A149; ML16071A150; ML16071A152; ML16071A153; ML16071A158; ML16071A160; ML16071A166; ML16071A170; ML16071A174; ML16071A176; ML16071A180; ML16071A181; ML16071A184; ML16071A185; ML16071A187; ML16071A188; ML16071A193; ML16071A195; ML16071A198; NL-15-1898, Enclosure 2: EAL Deviations and Differences Maxtrix (Farley, Hatch, and Vogtle) Part 1 of 4; NL-15-1898, Enclosure 2: EAL Deviations and Differences Maxtrix (Farley, Hatch, and Vogtle) Part 2 of 4; NL-15-1898, Enclosure 2: EAL Deviations and Differences Maxtrix (Farley, Hatch, and Vogtle) Part 3 of 4; NL-15-1898, Enclosure 2: EAL Deviations and Differences Maxtrix (Farley, Hatch, and Vogtle) Part 4 of 4; NL-15-1898, Enclosure 3: Hatch EAL Calculations - 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 1 of 3; NL-15-1898, Enclosure 3: Hatch EAL Calculations - 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 2 of 3; NL-15-1898, Enclosure 3: Hatch EAL Calculations - 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 3; 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 1 of 6; 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 2 of 6; 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; 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 4 of 6; 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 5 of 6; 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 6 of 6; NL-15-1898, Enclosure 5: Marked-Up EAL Schemes - 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 1 of 8; NL-15-1898, Enclosure 5: Marked-Up EAL Schemes - 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 2 of 8; NL-15-1898, Enclosure 5: Marked-Up EAL Schemes - 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 8; NL-15-1898, Enclosure 5: Marked-Up EAL Schemes - 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 4 of 8; NL-15-1898, Enclosure 5: Marked-Up EAL Schemes - 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 5 of 8; NL-15-1898, Enclosure 5: Marked-Up EAL Schemes - 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 6 of 8; NL-15-1898, Enclosure 5: Marked-Up EAL Schemes - 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 7 of 8; NL-15-1898, Enclosure 5: Marked-Up EAL Schemes - 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 8 of 8; NL-15-1898, Enclosure 5: Vogtle Marked-Up EAL Schemes - 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 1 of 4; ... further results
References
	NL-15-1898
	Download: ML16071A150 (44)
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V6 Page 1 of 6 SOUTHERN NUCLEAR DOCUMENT TYPE: PLANT E.I. HATCH ANNUNCIATOR RESPONSE PROCEDURE (ARP) AGE 1 OF 2 DOCUMENT 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/A ANNUNCIATOR RESPONSE PROCEDURES FOR 1H11-P654 -ALARM PANEL 1 ARP NO.VER NO.6.1 ARP NO.VER NO.ARP NO.VER. NO.SPARE 34AR-654-001

-1 34AR-654-031-1 4.0 34AR-654-061 34AR-654-002-1 1.0 34AR-654-032 SPARE 34AR-654-062-1 3.0 34AR-654-003-1 4.1 34AR-654-033-1 SPARE 34AR-654-063-1 2.0 34AR-654-004-1 6.1 34AR-654-034-1 2.0 34AR-654-064-1 3.0 34AR-654-005-1 3.0 34AR-654-035-1 2.0 34AR-654-065-1 7.3 34AR-654-006-1 7.1 34AR-654-036-1 1.0 34AR-654-066-1 2.0 34AR-654-007-1 12.0 34AR-654-037-1 4.0 34AR-654-067-1 8.0 34AR-654-008-1 4.0 34AR-654-038-1 4.0 34AR-654-068-1 8.0 34AR-654-009-1 5.2 34AR-654-039-1 2.0 34AR-654-069-1 2.0 34AR-654-0 10-1 2.1 34AR-654-040-1 2.0 34AR-654-070-1 3.1 34AR-654-01 1 SPARE 34AR-654-041-1 2.1 34AR-654-071-1 2.0 34AR-654-012-1 0.0 34AR-654-042-1 1.0 34AR-654-072-1 2.0 34AR-654-01 3-1 2.0 34AR-654-043-1 2.0 34AR-654-073-1 2.0 34AR-654-014-1 3.0 34AR-654-044-1 3.0 34AR-654-074-1 3.0 34AR-654-015 SPARE 34AR-654-045 SPARE 34AR-654-075-1 SPARE 34AR-654-0 16-1 4.0 34AR-654-046-1 3.0 34AR-654-076-1 3.0 34AR-654-017 SPARE 34AR-654-047-1 1.0 34AR-654-077-1 1.0 34AR-654-01 8-1 3.1 34AR-654-048-1 3.0 34AR-654-078-1 1.1 34AR-654-01 9-1 2.0 34AR-654-049-1 3.0 34AR-654-079-1 2.0 34AR-654-020-1 2.0 34AR-654-050-1 4.1 34AR-654-080-1 6.2 34AR-654-02 1-1 5.4 34AR-654-05 1-1 3.0 34AR-654-08 1-1 SPARE 34AR-654-022-1 11.1 34AR-654-052

-SPARE 34AR-654-082 SPARE 34AR-654-023-1 4.0 34AR-654-053

.SPARE 34AR-654-083 SPARE 34AR-654-024-1 2.0 34AR-654-054-1 3.0 34AR-654-084-1 2.0 34AR-654-025 SPARE 34AR-654-055-1 3.0 34AR-654-085-1 3.0 34AR-654-026 SPARE 34AR-654-056-1 1.1 34AR-654-086-1 2.0 34AR-654-027 SPARE 34AR-654-057-1 1.0 34AR-654-087-1 2.0 34AR-654-028-1 1.0 34AR-654-058-1 2.0 34AR-654-088-1 2.0 34AR-654-029-1 2.0 34AR-654-059-1 3.0 34AR-654-089-1 3.0 34AR-654-030 SPARE 34AR-654-060-1 0.0 34AR-654-090-1 SPARE NOTE. 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. j I Level Of Use ARPs CONTINUOUS ALL REFERENCE None INFO None NMP-AP-002 V6 Page 2 of 6 PAGE 2 OF2 SOUTHERN NUCLEAR PLANT E.I. HATCH DOCUMENT TITLE: DOCUMENT NUMBER: ARP'S FOR CONTROL PANEL 1HI11-P654,I 34AR-654-901-1 ALARM PANEL 1 VERSION NO: 23.0 UNIT 1 1 H11-P654 -1 (Left)654-00 1 654-002 654-003 654-004 654-005 654-006 654-007 654-008 654-016 654-017 654-018 654-019 654-020 654-021 654-022 654-023 654-03 1 654-032 654-033 654-034 654-035 654-036 654-037 654-038 654-046 654-047 654-048 654-049 654-050 654-05 1 654-052 654-053 654-06 1 654-062 654-063 654-064 654-065 654-066 654-067 654-068 654-076 654-077 654-078 654-079 654-080 654-08 1 654-082 654-083 UNIT I 1H11-P654

-1 (Right)654-009 654-010 654-011 654-012 654-013 654-014 654-015 654-024 654-025 654-026 654-027 654-028 654-029 654-030 654-039 654-040 654-04 1 654-042 654-043 654-044 654-045 654-054 654-055 654-056 654-057 654-058 654-059 654-060 654-069 654-070 654-07 1 654-072 654-073 654-074 654-075 654-084 654-085 654-086 654-087 654-088 654-089 654-090 NMP-AP-002 V6 Page 3 of 6 1.0 IDENTIFICATION:

ALARM PANEL 654 SPENT FUEL STORAGE POOL LEVEL LOW DEVICE: 1G41-N362 Level Sensor 1G41-N372 Remote Electronics SETPOINT: 225 ft. 9 in.(21' -7" above the top of the fuel assemblies seated in the Fuel Pool)5.0 OPERATOR ACTIONS:* 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.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 damage the DHR pump.5.1 Enter 34AB-G41-002-1, Decreasing Rx Well/Fuel Pool Water Level. Li 5.2 IF Fuel Pool gates are installed:

5.2.1 Raise water level by regulating 1G41-F041, Spent Fuel Pool Make-up water from CST Valve, located at 185RBR07, panel 1H21-P155.

LI 5.2.2 Confirm Fuel Pool Cooling filter effluent is returning to fuel pool per 34SO-G41-003-1, Fuel Pool Cooling and Cleanup System. LI 5.3 IF Fuel Pool gates are NOT installed, request Maintenance to install gates. LI 5.4 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. LI 6.0 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 3.7.8 7.2 H-17074, Fuel Pool Cooling System G41 Elem Diag 34AR-654-022-I VER. 11.1 MGR-0048 Ver. 5.0NMPA02 NMP-AP-002 V6 Page 4 of 6 DOCUMNT TTLE:DOCUMENT NUMBER: VERSION NO: ARP'SALRFOR CONTROLpAE PANEL 2H11-P654, 34AR-654-901-2 23.4 EXPI RATION APPROVALS:

EFFECTIVE DATE: DEPARTMENT MANAGER DAK for KDL DATE 09-07-12 DATE: 2-27-15 N/A SSM /PM N/A DATE N/A _____ANNUNCIATOR RESPONSE PROCEDURES FOR 2H11-P654

-ALARM PANEL 1 ARP 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 SPARE 34AR-654-002-2 2.1 34AR-654-032-2 2.1 34AR-654-062-2 3.1 34AR-654-003-2 3.1 34AR-654-033 SPARE 34AR-654-063 SPARE 34AR-654-004-2 2.2 34AR-654-034-2 2.3 34AR-654-064-2 2.3 34AR-654-005-2 2.0 34AR-654-035-2 2.1 34AR-654-065-2 4.2 34AR-654-006-2 8.0 34AR-654-036-2 3.4 34AR-654-066-2 5.1 34AR-654-007-2 11.0 34AR-654-037-2 2.0 34AR-654-067-2 6.0 34AR-654-008-2 3.1 34AR-654-038-2 4.1 34AR-654-068-2 6.0 34AR-654-009 SPARE 34AR-654-039 SPARE 34AR-654-069 SPARE 34AR-654-01 0 SPARE 34AR-654-040-2 3.0 34AR-654-070-2 4.2 34AR-654-01 1 SPARE 34AR-654-041-2 3.1 34AR-654-071-2 4.1 34AR-654-012-2 3.2 34AR-654-042-2 2.1 34AR-654-072-2 4.1 34AR-654-013-2 2.2 34AR-654-043-2 2.1 34AR-654-073-2 4.1 34AR-654-014-2 1.2 34AR-654-044-2 2.1 34AR-654-074-2 5.4 34AR-654-015-2 3.2 34AR-654-045 SPARE 34AR-654-075 SPARE 34AR-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 2.2 34AR-654-020-2 3.2 34AR-654-050 SPARE 34AR-654-080-2 7.5 34AR-654-021-2 7.1 34AR-654-051 SPARE 34AR-654-081 SPARE 34AR-654-022-2 11.1 34AR-654-052 SPARE 34AR-654-082 SPARE 34AR-654-023-2 6.1 34AR-654-053 SPARE 34AR-654-083 SPARE 34AR-654-024 SPARE 34AR-654-054 SPARE 34AR-654-084 SPARE 34AR-654-025 SPARE 34AR-654-055-2 3.0 34AR-654-085-2 4.2 34AR-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 4.1 34AR-654-029 SPARE 34AR-654-059-2 2.2 34AR-654-089 SPARE 34AR-654-030 SPARE 34AR-654-060 SPARE 34AR-654-090 SPARE I I I INOTQE: I Approval signature on this page constitutes approval for all procedures listed above at the Iversion indicated.

Tab numbers in the back correspond to procedure sequence number.I lit rm Level Of Use ARPs CONTINUOUS ALL REFERENCE None INFO None V6 Page 5 of 6 2H11-P654-1 (Left)654-001 654-002 654-003 654-004 654-005 654-006 654-007 654-008 654-016 654-017 654-018 654-019 654-020 654-021 654-022 654-023 654-031 654-032 654-033 654-034 654-035 654-036 654-037 654-038 654-046 654-047 654-048 654-049 654-050 654-051 654-052 654-053 654-061 654-062 654-063 654-064 654-065 654-066 654-067 654-068 654-076 654-077 654-078 654-079 654-080 654-081 654-082 654-083 UNIT 2 2H11-P654

-1 (Right)654-009 654-010 654-011 654-012 654-013 654-014 654-015 654-024 654-025 654-026 654-027 654-028 654-029 654-030 654-039 654-040 654-041 654-042 654-043 654-044 654-045 654-054 654-055 654-056 654-057 654-058 654-059 654-060 654-069 654-070 654-071 654-072 654-073 654-074 654-075 654-084 654-085 654-086 654-087 654-088 654-089 654-090 V6 Page 6 of 6 1.0 IDENTIFICATION:

ALARM PANEL 654 ________" --t---SPENT FUEL STORAGE POOL LEVEL LOW DEVICE: SETPOINT: 2G41-N372 Remote Electronics 226' 2.5" (approx. 22' 0.5" above the top of 2G41-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. Li 5.2 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.

LI 5.2.2 Confirm the Fuel Pool Cooling Filter effluent is returning to Fuel Pool per 34S0-G41-003-2, Fuel Pool Cooling and Cleanup System. Li 5.2.3 Confirm air supply pressure to seals is between 33 PSIG AND 37 PSIG AND 2P51-F549, 2P51-F563, and 1P51-F555 (unit one's air supply valve), Air Supply Valves, are OPEN. Li 5.3 IF Fuel Pool gates are NOT INSTALLED request maintenance to INSTALL gates. Li 5.4 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. Li 6.0 CAUSES: 6.1 Water loss from normal evaporation 6.5 Reactor well leakage 6.2 Fuel Pool liner leakage 6.6 Dryer/Separator storage pooi leakage 6.3 Fuel Pool gate leakage 6.7 Fuel Pool to transfer canal gate leakage 6.4 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 Storage 7.2 H-27736, Fuel Pool Cooling System Elem Pool Water Level 34AR-654-022-2 VER. 11.1 V7, Page 1 of 3 Fission Product Barrier Emerqency Action Levels Fuel Clad Barrier: Emergency Action Levels Fuel Clad Barrier Potential Loss Threshold 2.A RPV water level cannot be restored and maintained above -155 inches or cannot be 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 fuel was increased from 148" to 150" inches. The Appendix A is attached below. Thus the 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 3 Fuel Clad Barrier Loss Threshold 4.A DWRRM greater than 1,400 R/hr.In Attachment C the detector radiation level of 1.4E3 R/hr was calculated.

The calculation used core inventory from NL-06-1637 to calculate isotopes concentrations.

The calculation for DEI131 was performed to find a ratio to DEl 300uCi/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.A Offgas Pre- and Post-Treatment Monitors Offscale High AND Fission Product Monitor 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 product monitors D11IP010 (section B of Attachment D). It was found that these instruments will be off scale.RCS Barrier: Emergency Action Levels RCS Barrier Loss Threshold I .A Primary containment pressure greater than 1.85 psig due to RCS leakage.LIS 1C71N650A-D 1.85 psig LIS 2C71N650A-D 1.85 psig References (H 16568, PDMS)Ret PDMS Ref. PDMS RCS Barrier Loss Threshold 2.A RPV water level cannot be restored and maintained above -155 inches or cannot be 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.A DWRRM greater than 40 R/hr.In Attachment E the detector radiation level of 40 R/hr was calculated.

The calculation used core inventory from NL-06-1637 to calculate isotopes concentrations.

The calculation for DE/13I was performed to find a ratio to DEl V7 Page 3 of 3 Table 1 -REACTOR VESSEL LEVEL SETPOINTS AND ACTIONS Setpoints*

+54.0 (+54.5")+51.7"+42"+37"+/-32"<+30 ->+/-+20+3"-35"-60-101"-155-193" Action Trips MAIN AND RFP Turbines Trips HPCI and RCIC High Level alarm from Level Recorder R608 Normal Operating Level Low Level alarm from Level Recorder R608, input to Recirculation Pump Runback to SL #2 on a loss of a RFP Input to Variable Recirculation Pump Runback to SL #4 Reactor Scram, PCIS Group II, ADS Permissive (Confirmatory), Closes Shutdown Cooling Isolation Valves Start HPCI, RCIC and SBGT, Isolates Secondary Containment (R/F and R/B zones), PCIS Group V (RWCU), ARI Initiates Trips Reactor Recirculation Pumps PCIS Group I (MSIVs), Start Core Spray, LPCI and Diesel Generators, ADS Permissive, Trips CRD pumps, Closes PSW isolation to Turbine Bldg, Control Room Ventilation switches to Pressurization Mode Top of Active Fuel Containment Spray Permissive (2/3 core coverage)*Referenced to instrument zero.

V8, Page 1 of 3 CA1 Loss of RPV Inventory.

Operability Mode Applicability:

Emergency Actuation Levels Cold 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 set point. According to H16 145 and H26189 the level 2 instruments are 1/2D21N692A-D and 1/2D21N682A-D.

According to PDMS the instruments are set for -35g.2.a. RPV level cannot be monitored for 15 minutes or longer AND b. UNPLANNED level increase in any of the following due to a loss of RPV inventory.

Drywell Floor Drain Sumps Drywell Equipment drain Sumps Torus Torus room Sumps Reactor Building Floor Drain Sumps Turbine Building Floor Drain Sumps Rad Waste Tanks V8, Page 2 of 3 CS1 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.

AND b. RPV level less than -41" (6" below the Level 2 actuation setpoinl According to S25213 page 12 and 14 ECCS actuates at level 2 acth According 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 and 5". Therefore

-35" a. Secondary CONTAINMENT INTEGRITY established.

AND b. RPV level less than -158" (TOAF).The reactor vessel top of active fuel was calculated in Fuel Clad Barrier Potential 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 Sumps Drywell Equipment Drain Sumps Turbine Building Floor Drain Sumps Torus Rad Waste Tanks Torus Room Sumps V8 Page 3 of 3 Table 1 -REACTOR VESSEL LEVEL SETPOINTS AND ACTIONS Setpoints*

+54.0 (+/-54.5")+/-51.7"+/-42"+37"+32"<+30 -> +20+/-3" Action Trips MAIN AND RFP Turbines Trips HPCI and RCIC High Level alarm from Level Recorder R608 Normal Operating Level Low Level alarm from Level Recorder R608, input to Recirculation Pump Runback to SL #2 on a loss of a RFP Input to Variable Recirculation Pump Runback to SL #4 Reactor Scram, PCIS Group II, ADS Permissive (Confirmatory), Closes Shutdown Cooling Isolation Valves Start HPCI, RCIC and SBGT, Isolates Secondary Containment (R/F and R/B zones), PCIS Group V (RWCU), ARI Initiates Trips Reactor Recirculation Pumps PCIS Group I (MSIVs), Start Core Spray, LPCI and Diesel Generators, ADS Permissive, Trips CRD pumps, Closes PSW isolation to Turbine Bldg, Control Room Ventilation switches to Pressurization Mode Top of Active Fuel Containment Spray Permissive (2/3 core coverage)-60-101"-155-193"*Referenced to instrument zero.

V9, Page 1 of 3 CA1 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 set point. According to H16 145 and H26 189 the level 2 instruments are 1/2D21N692A-D and 1/2D21N682A-D.

According to PDMS the instruments are set for -35".2.a. RPV level cannot be monitored for 15 minutes or longer AND b. UNPLANNED level increase in any of the following due to a loss of RPV inventory.

Drywell Floor Drain Sumps Drywell Equipment drain Sumps Torus Torus room Sumps Reactor Building Floor Drain Sumps Turbine Building Floor Drain Sumps Rad Waste Tanks V9, Page 2of 3 CS1 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.

AND b. 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 and 1/2D21N682A-D.

According to PDMS the instruments are set for -35". Therefore

-35"-6" =-41" 2.a. Secondary CONTAINMENT INTEGRITY established.

AND b. RPV level less than -158" (TOAF).The reactor vessel top of active fuel was calculated in Fuel Clad Barrier Potential 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 Sumps Drywell Equipment Drain Sumps Turbine Building Floor Drain Sumps Torus Rad Waste Tanks Torus Room Sumps V9, Page 3of 3 CG1 Loss of RPV Inventory Affecting Fuel Clad Integrity with Containment Challenged.

Operating Mode Applicability:

Emergency Action Level: Cold Shutdown Refueling (1 OR 2)1.a. RPV level less than -158" (TOAF)The reactor vessel top of active fuel was calculated in Fuel Clad Barrier Potential Loss 2.A. The top of fuel was found to be at vessel level of -158.44".This can be rounded to -158" AND b. ANY indication from the Containment Challenge Table C1.2.a. RPV level cannot be monitored for 30 minutes or longer AND 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 Sumps Drywell Equipment Drain Sumps Turbine Building Floor Drain Sumps Torus Rad Waste Tanks Torus Room Sumps Radiation monitor readings indicative of core unco very are investigated in Attachments J and K resulting in no monitors able to provide on-scale indications of core uncovery.AND c. ANY indication from the Containment Challenge Table Cl Containment Challenge Table C1 Containment I-t12 greater than or equal to 6% AND 02 greater than or equal to 5%Primary Containment Pressure:

greater than 56 psig Secondary 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 not required.Damage to a loaded cask CONFINEMENT BOUNDARY E-HUI VlO Page 1 of 5 Containment and Radioactivity Release Control 8WROG EPGs'SAGs, Appendix B EPG/SAG Step (PCIG, continued)

Control hydrogen and oxygen concentrations in the dr~swell as follows:..........

II Ill II Illllll Illll Ill I C IiI cennot be dotermined to be below 5%Sugmua Chauber Hyirogm Ce kncela< 5%Nan.Disdsd

'6%a 6%t or ceqc bedlm *,e 6.t Is h&em a* o~Iet No action No scion tot be eow 6 _ _ _ _ __ _ _ _ _ _ _Control hydrogen and oxygen concentrations in the suppression chamber as follows: Supmseo Calmber OzygenCwuontue~on a 5% or cannot be detem~ined to be below 5%OwnHl e m< 5%NaweOeidid a 86% ot owwwt SO be bedsw 6 SNoracto No action N__,____ reqired en ,UE aO%orcennot

be deterine Revision 3 Revision 3 Mark l and)! containments only2-61 B-16-21 VlO Page 2 of 5 BWROG EPGs'SAGs, Appendix B Containment and Radioactivity Release Control Discussion (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 be determined) and the hydrogen concentration in either volume is equal to or greater than 6% or cannot be determined.

Under these conditions, a potential for deflagration exists. Venting is then permitted irrespective of the resulting radioactivity release rate and the purge may be performed using either air or nitrogen.

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 is actually detected or if the concentration cannot be determined.

The specified concentrations of 6% hydrogen and 5% oxygen are the minimum values that 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.

The resulting pressure peak may be high enough to rupture the primary containment or damage the drywell-to-wetwell boundary.Note that when oxygen concentration in one volume is equal to or greater than 5% or cannot 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 from the other volume could migrate to the deinerted area, creating a deflagrable mixture before 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 be above the value required to support combustion.

The branch tables therefore specify steps to perform if hydrogen or oxygen concentration cannot be determined relative to its deflagration limit. Failure or unavailability of the normal monitor, however, does not necessarily mean that a gas concentration cannot be determined.

The containment is normally inerted and hydrogen generation rates are expected to be relatively slow. If the most recent data showed considerable margin to the deflagration limits and conditions have not changed significantly since the readings were taken, it is thus unnecessary to assume that the concentrations immediately exceed the limits when direct measurement capability is lost. Rather, a decision that hydrogen and oxygen concentrations cannot be determined requires a judgment considering plant conditions, parameter trends, and the availability of alternate indications.

Revision 3 Revision 5 Mark Iland 11 containmenhs only 1-2 B.16-23 Vl0, Page 3of 5 CG1 Loss of RPV Inventory Affecting Fuel Clad Integrity with Containment Challenged.

Operating Mode Applicability:

Emergency Action Level: Cold Shutdown Refueling (1 OR 2)1.a. RPV level less than -158" (TOAF)The reactor vessel top of active fuel was calculated in Fuel Clad Barrier Potential Loss 2.A. The top of fuel was found to be at vessel level of -158.44".This can be rounded to -158" AND b. ANY indication from the Containment Challenge Table C1.2.a. RPV level cannot be monitored for 30 minutes or longer AND 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 Sumps Drywell Equipment Drain Sumps Turbine Building Floor Drain Sumps Torus Rad Waste Tanks Torus Room Sumps Radiation monitor readings indicative of core unco very are investigated in Attachments J and K resulting in no monitors able to provide on-scale indications of core unco very.AND c. ANY indication from the Containment Challenge Table Cl~Containment Challenge Table C1 Containment I-2 greater than or equal to 6% AND 02 greater than or equal to 5%Primary Containment Pressure:

greater than 56 psig Secondary 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 not required.Damage to a loaded cask CONFINEMENT BOUNDARY E-HU1 Vl0, Page 4of 5 2uCiigm. GRODEC was used to calculate the fluence within the drywell. Cylinder geometry was used to calculate the geometric fraction.RCS Barrier Loss Threshold 5.A Drywell Fission Product Monitor reading 5.0 x 105 cpm.This EAL is to cover drywell fission product monitor indications that may indicate loss or potential loss of the RCS barrier. Attachment F determined that the reading on drywell fission product monitor D1 1K630 of 1E6 cpm will indicate potential loss of RCS barrier. Per SX18062 page 34 the monitor K630 range is 10 to 10^6 cpm.Primary Containment Barrier: Emergency Action Level: Primary Containment Barrier Potential Loss Threshold I .A Primary 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 .B Greater 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.A DWRRM greater than 26,000 R/hr.The evaluation of expected radiation readings on DWRRM (DI11K621) was performed in Attachment G of this calculation.

The detector is expected to read 2. 6E4R/hr. The range of this instrument is 1-10^7 R/hr (established in attachment C).

V10, Page 5of 5 27. H 16568 V5.0 "REACTOR PROTECTION SYSTEM P&ID" 28. BH2-M-V999-.0047 V2.0 "DRYWELL EQUIPMENT EQ DOSES FOR EXTENDED POWER UPRATE FOR REA HT-96660" 29. HNP Technical Specifications 273/218 01-07-16 30. NUREG-0016 "Calculation of releases of radioactive materials in gaseous an liquid effluents from boiling water reactors (BWR GALE Code)" April 1976.31. RG 1.183 "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors." July 2000 32. 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 GRODEC COMPUTER PROGRAM" 37. Deleted 38. 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 & ID AND 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" Vll Page 1 of 3 IHNP-1-FSAR-5 In the event of a process system piping failure within the drywell, reactor water and steam are released into the drywetl gas space. The resulting increased drywell pressure forces a mixture of air, steam, and water through the vent system into the suppression pool. The steam condenses rapidly in the suppression pool, resulting in rapid pressure reduction in the drywell.Air transferred during reactor blowdown to the suppression chamber pressurizes the chamber and is subsequently vented to the drywell through the vacuum relief system as the pressure in the drywell drops below that in the suppression chamber. Cooling systems remove heat from the drywell and the suppression pool for continuous cooling of the primary containment under the postulated DBA conditions.

Isolation valves ensure the containment of radioactive material within 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 Its design parameters during normal operation.

The primary containment system design loading considerations are provided In chapter 12 and appendix K. The safety analysis presented In HNP-2-FSAR chapter 15 demonstrates the effectiveness of the primary containment system as a 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 Drwl The drywell is a steel pressure vessel with a spherical lower portion 65 ft in diameter and a cylindrical upper portion 35 ft 7 in. in diameter.

The overall height of the drywell Is ~111 ft. The design, fabrication, inspection, and testing of the drwell comply with the requirements of the ASME Code,Section III, Subsection B, Requirements for Class B Vessels, which pertains to containment vessels for nuclear power stations.

The primary containment is fabricated of SA-516 grade 70 plates.IThe drywell Is designed for an internal pressure of 56 psig lcincident with a temperature of 281"F, with applicable dlead, live, and seismic loads imposed on the shell. Thermal stresses in the steel shell due to temperature gradients are also Incorporated into the design. Thus, in accordance 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 of the steel drywell shell. Charpy V-notch specimens were used for impact testing of plate and forging material to verify proper material properties.

Plates, forgings, and pipe associated with the drywell have an Initial nil ductility transition temperature (NDTT) < 0°F when tested In accordance with the appropriate code for the materials.

The drywell Is assumed to be neither pressurized nor subjected to substantial stress at temperatures below 30°F.The drywell is enclosed in a reinforced concrete structure for shielding purposes.

Resistance to deformation and buckling of the drywell plate Is provided In areas where the concrete backs up the steel shell. Above the transition zone, the drywell Is separated from the reinforced concrete by 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 up to 38 ft in two separate layers of 3 segments, each weighing 180 klps. The plug segments are designed for 1000 Iblft uniform floor loading and were checked for the effects of the tornado C 5.2-3 5.2-3REV 31 9113 Vll Page 2 of 3 SHNP-1FSR-5 TABLE 5.2-7 PRIMARY CONTAINMENT SYSTEM DESIGN PARAMETERS General Information Design Pressure Internal -dryweill~6Op1

-- -suppression chamber 58.0pi-suppression chamber Design Temperature Drywell Suppression chamber Free Volume Drywell (including vent system)Suppression chamber-approximate minimum-approximate maximum Leakage Rate Downicomer Submergence Overall Vent Resistance Loss Factor Pool Depth (Normal)No. of Vents Normal Vent Diameter (ID)Total Vent Area No. of Downcomers Nominal Downcomer Diameter 2.0 psig 2.0 psig 281°F 146,010 ft 112,900 ft 115,900 ft 3 1.2% free vol/day 4 ft 0 in.(axb)4.4(0), (5.5 1)CaXb)12ft4 In.8 5ftl 11In.220 ft 2 80 2.0 ft I I a. 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 V11 Page 3 of 3 IN--FA-1, Drywell and Vent Systems Design internal pressure 't3QF Operating internal pressure < 2 psig at 150°F 2. Suppression Chamber Design Internal pressure 56 psig at 340°F Operating intemnal pressure < 2 psig at 50° to 100°F The design internal pressure is 90% of the maximum internal pressure.Pipe Rupture Loads (Yr, Y 1 , Yin)Yr= Equivalent static load on the structure generated by the reaction on the broken high-energy pipe during the postulated break, and including an appropriate dynamic load factor to account for the dynamic nature of the load.Yj= Jet impingement equivalent static load on a structure generated by the postulated break and including an appropriate dynamic load factor to account for the dynamic nature of the load.The containment is designed for the following jet impingement loads resulting from pipe ruptures within the containment:

Area of Location Jet ForceInuee Drywell sphere 709,000 lb 3.94ft Duywell knuckle 472,000 lb 2.63 ft 2 Drywell cylinder up to el 203 ft 9 in. 472,000 lb 2.63 ft 2 Drywell head 32,600 lb 0.181ft The jet forces consist of steam and/or water at 340°F. Only one of the above 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 during the 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 the event of a LOCA, the entire containment, including the suppression chamber, vent system, and the drywell, are flooded up to el 227 if, and the resulting hydrostatic load, FL, was considered In the containment design.3.8-11 3.8-11REV 31 9/13 V12 Page 1 of 6 PAGE 14 OF 30 SOUTHERN NUCLEAR PLANT E.I. HATCH DOCMEN TTLE IDOCUMENT NUMBER: VERSION NO: SECONDARY CONTAINMENT CONTROL 34AI3-T22-003-1 5.14 ATTACHMENT 6 ATTACHMENT PAGE: TITLE: SECONDARY CONTAINMENT OPERATING RADIATION LEVELS I OF 3 HVAC EXHAUST RADIATION MAXIMUM NORMAL OPERATING ANNUNCIATORS ON 1H11-P601 VALUE mr/hr HI-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) 1 Max Normal Max Safe AREA RADIATION MONITORS Operating Operating on 1H11-P600, 1D21-P600 Value Value mR/hr mR/hr REFUEL FLOOR AREA 1 Reactor head laydown area (1 D21-K601A) 50 1000 2 Refueling Floor Stairway (1 D21-K601 B) 50 1000 3 Refueling Floor (1D21-K601D) 50 1000 4 Drywell Shield Plug (1D21-K601E) 50 1000 5 Spent Fuel Pool & New Fuel Storage (1D21-K601M) 50 1000 203' ELEVATION AREA 6 RB 203' Working Area (1D21-K601X) 50 1000 185' ELEVATION AREA 7 Spent Fuel Pool Demin. Equip (1D21-K601C) 150 1000 8 Fuel Pool Demain. Panel (1D21-K617) 50 100 158' ELEVATION AREA 9 RB 158' Working Area (1D21-K601 K) 50 1000 10 Rx Wtr Sample Rack Area 158' (1 D21-K601 L) 50 1000 130' ELEVATION NORTH AREA 11 TIP Area (1D21-K601F) 50 1000 12 North CRD HCU (1D21-K6O1P) 50 1000 13 TIP Probe Drives Area (1D21-K601 U) 100 1000 MGR-0009 Rev 5.0 Vi12 Page 2 of 6 PAGE 15 OF 30 SOUTHERN NUCLEARI PLANT E.I. HATCH _DOUMNTTILE tDOCUMENT NUMBER: VERSION NO: SECONDARY CONTAINMENT CONTROL 1 34AB-T22-003-1 5.14 ATTACHMENT 6 ATTACHMENT PAGE: TITLE: SECONDARY CONTAINMENT OPERATING RADIATION LEVELS 2 OF 3 Max Normal Max Safe AREA RADIATION MONITORS Operating Operating on 1 H 11 -P600, 1 D2 1-P600 Value Value mR/hr mR/hr 130' ELEVATION EAST AREA 14 RB 130' N-E Working Area (1D21-K601G) 50 1000 15 Equipment Access Airlock (1D21-K601S) 50 1000 130' ELEVATION SOUTH AREA 16 RB 130' S-W Working Area (1 D21-K601 H) 50 1000 17 South CRD HCU (1D21-K601N) 50 1000 SOUTHWEST DIAGONAL AREA 18 RCIC Equip S-W Diagonal (1 D21-K601V) 50 1000 NORTHWEST DIAGONAL AREA 19 CRD Pump N-W Diagonal (1D21-K601W) 50 1000 NORTHEAST DIAGONAL AREA 20 CS & RHR N-E Diagonal (1D21-K601Y) 50 1000 NORTHEAST DIAGONAL AREA 21 CS & RHR S-E Diagonal (1D21-K601R) 50 1000 HPCI AREA 22 HPCl turbine Area (1D21-K601T) 150 1000 Detector to Trip Unit cross reference DETECTOR TRIP UNIT iDli-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-K608A 1 DI1-N017A

&I1D11-N017B 1ID11-K616A

& 1D11-K616B MGR-0009 Rev 5.0 Vi 2 Page 3 of 6 PLANT E.I. ATCHPAGE 29 OF 30 DOUMNTTILE IDOCUMENT NUMBER: VERSION NO: SECONDARY CONTAINMENT CONTROL I 34AB-T22-O03-1 5.14 ATTACHMENT 10 ATTACHMENT PAGE: TITLE: SECONDARY CONTAI NMENT PARAMETERS 1 OF 1 V12 Page 3 of 6 101 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 AREA V~x 150 67 180 67 150 67¶2125 t6 212.5 98 212.5 98 14" eowsr" ev.15 atxm 67' eI.~.$0U11df.~ST W)AGONAL AREA 4" akowe 3? dzow 5r7' ete',r SO)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 150 67 150 67 150 67 212.5 98 212.5 98 212.5 98 on 1HtI-PUO, 1D21:.P6 ____ AREA 2 RueabFloorS iy(10214(501B) 50 3 Refje&~iwr9

~(1D21.K601D) 50 1000 4 D~wvul Shield Pbig 1D214(601E) 50 1O000 5 50 1000 203 ELEVATIO AREA '6 R8 203" 50 1O00 7 Spent Fuel PVCoo Oe Equip (1D21.K601 C) 150 1000 8 Fuel Pod Dem, Paule(1D21,,K617) 50 100 _9 RB1I Wdd 50 1Q00 10 D21.K60L) 50 1O000 ___11 ELEVATION NOrTH ARE 11 TIP ,,are(102t-KOI F) 50 1O000 I2 NohCO HU (1O2l..K60lP) 50 1000 13 TIPv Po Dvee Area (tD2I-KB0t U) !0 10 ___0 14 510 N.EVATodN EArT (AREA60G) 5 1 14 Equipent/:¢rskudd(1D214(601S) 50 1000 1Eqmft30ss.EATIx*(1D2.Km 1S 50 100 16 510' E. WVlorkngSU Area (01K0H 0 10 17 Soth CRD HCU(1D21-KE01N) 50 1ooo SOUTHWAST DIAGONAL,,REA 18 RCICEcp~ SW Diagn(1O 2t.4( )'1V) 50 1000 NORTHWEST DIPGG-NAL REA- -- -19 RD Purp W DIacxaa(1D214(8OlW]

50 1000 NOGll'*AST D AREA- -20 C & RIR N.qEneol (1D21.K501Y) 50 1000-O~ES DIAGONAJ NaARE 21 CS&mRffS.O~eig(t1O21.W98R) 50 1O000 150' ELEVATION (OT~ET 22 IPCITwIoie/rea (10214011)l) 150 1000 Wi-'S '--l' I =LM A 119 Irh ShnTel ltB21.N14) 195 30¶20 01FF I6 5 OPS-1933Vr4.

34AB-T22-O03-1 OPS-1 933 MGR-0009 Rev 5.0 Vi12 Page 4 of 6 PAGE 14 OF 37 SOUTHERN NUCLEAR PLANT E.I. HATCH DOCMEN TTLE IDOCUMENT NUMBER: VERSION NO: SECONDARY CONTAI NM ENT CONTROL 34AB-T22-003-2 4.2 ATTACHMENT 6 ATTACHMENT PAGE: TITLE: SECONDARY CONTAINMENT OPERATING RADIATION LEVELS 1 OF 4 MAX NORMAL HVAC EXHAUST RADIATION ANNUNCIATORS OPERATING ON 2H1 1-P601 VALUE mR/hr HI-HI REACTOR BUILDING (RX) RADIATION ANNUNCIATOR

-RX BLDG POT CONTAM AREA RADIATION (2D1 1-K609 A-D) 18 REFUELING 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) 5.7 MAX NORMAL MAX SAFE AREA RADIATION MONITORS OPERATING OPERATING ON 2H1 1-P600, 2D21-P600 VALUE VALUE mR/hr mR/hr REFUEL FLOOR AREA 1. Reactor head laydown area (2D21-K601A) 50 1000 2. Dryer separator pool (2D21-K601 E) 50 1000 3. Spent Fuel Pool & New Fuel Storage (2D21-K601M) 50 1000 4. Reactor Vessel Refueling Floor (2D21-K61 1K) 50 1000 5. Reactor Vessel Refueling Floor (2D21-K611 L) 50 1000 203' ELEVATION AREA (EAST)6. CRD repair area (2D21-K601T) 50 1000 203' ELEVATION AREA (WEST)7. HVAC Room West El. 203' (2D21-K600D) 50 100 Vi12 Page 5 of 6 PAGE 15 OF 37 SOUTHERN NUCLEAR PLANT E.I. HATCH DOCMEN TTLE IDOCUMENT NUMBER: VERSION NO: SECONDARY CONTAI NM ENT CONTROL 34AB-T22-003-2 4.2 ATTACHMENT 6 ATTACHMENT PAGE: TITLE: SECONDARY CONTAINMENT OPERATING RADIATION LEVELS 2 OF 4 MAX NORMAL MAX SAFE AREA RADIATION MONITORS OPERATING OPERATING ON 2H11-P600, 2D21-P600 VALUE VALUE____________________________________

mR/hr mR/hr 185' ELEVATION AREA 8. Spent fuel pool passageway (2D21-K601 P) 50 1000 9. RB 185' operating floor (2D21-K601R) 50 1000 10. RB 185' sample panel area (2D21-K601S) 50 1000 11. RB 185' RWCU control panel (2D21-K601U) 150 1000 158' ELEVATION AREA (NORTH)12. RB 158' area N-E (2D21-K601C) 50 1000 13. RB 158' area N-W (2D21-K601D) 50 1000 158' ELEVATION AREA (SOUTH)14. RB 158' area S-E (2D21-K601B) 50 1000 15. Decant pump & equipment room area 158' (2D21-K601 L) 50 1000 130' ELEVATION AREA (NORTHWEST)

16. Tip area (2D21-K601 F) 50 1000 130' ELEVATION AREA (NORTHEAST)

17. RB 130' N-E working area (2D21-K601G) 50 1000 130' ELEVATION AREA (SOUTHEAST)

18. South CRD HCU (2D21-K601N) 50 1000 130' ELEVATION AREA (SOUTHWEST)

19. RB 130' S-W working area (2D21-K601 H) 50 1000 NORTHWEST DIAGONAL AREA 20. RCIC equipment N-W diagonal (2D21-K601V) 50 1000 SOUTHWEST DIAGONAL AREA 21. CRD pump S-W diagonal (2D21-K601W) 50 1000 NORTHEAST DIAGONAL AREA 22. CS & RHR N-E diagonal (2D21-K601X) 50 1000 SOUTHEAST DIAGONAL AREA 23. CS & RHR S-E diagonal (2D21-K601Y) 150 1000 Vi12 Page 6 of 6 SOUTHERN NUCLEAR PLANT E.I. HATCH PAGE 35 OF 37 DOCMEN TILE:DOCUMENT NUMBER: VERSION NO: SECONDARY CONTAINMENT CONTROL 34AB-T22-003-2 4.2 ATTACHMENT 1"1 ATTACHMENT PAGE: TITLE: SECONDARY CONTAINMENT PARAMETERS 1 OF 1 I ( **t]%'I* ____T 113 212,5 114$ W-(2E51-N05O) 116 TorusVe 42 I e~- ~t*I~a' g CUAIVWJ A~~A 1UAIV'1 I~6 CD (021K60T)50 1000 203 AREA (WT)7 HVAC RO(XWest O3'{2O21.M0GO) 50 100 a Sp~tFug oo!Pa0 gway(2S.IR

1) 50 10(00 11RB1WR' 150 1000 10 Swll (20214( 601D ) 50 1000)tt 8 1' 50 1000 130 ELEVATION AR-A (NORTFAS)12 R51U8 raN-E k2et 4(01C) 50 1000 18 R 15UhCHC 50 1Q000 130 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 AREA 221CRPupS.WdagouI(22141601W) 50 1000(2D21.K601Y) 150 1000: 4A- 1I2-003-2 OPS-1 932 OPS-1932 Vi13 Page 1 of 5 AC Sources -Operating B 3.8.1 B 3.8 ELECTRICAL POWER SYSTEMS B 3.8.1 AC Sources -Operating BASES BACKGROUND The Unit 1 Class 1 E AC Electrical Power Distribution System AC sources consist of the offsite power sources (preferred power sources, 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 power system provides independence and redundancy to ensure an available source of power to the Engineered Safety Feature (ESF)systems.The Class I1E AC distribution system is divided into redundant load groups, so loss of any one group does not prevent the minimum safety functions from being performed.

Each load group has connections to two preferred offsite power supplies and a single 0G.Offsite power is supplied to the 230 kV and 500 kV switchyards from the transmission network by eight transmission lines. From the 230 kV switchyards, two electrically and physically separated circuits provide AC power, through startup auxiliary transformers 1C and ID, to 4.16 kV ESF buses 1E, 1F, and 1G. A detailed description of the offsite power network and circuits to the onsite Class 1 E ESF buses is found 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 power from the offsite transmission network to the onsite Class I1E ESF bus or buses.Startup auxiliary transformer (SAT) 1D provides the normal source of power to the ESF buses 1E, 1iF, and 1G. If any 4.16 kV ESF bus loses power, an automatic transfer from SAT ID to SAT IC occurs.At this time, 4.16 kV buses IA and lB and supply breakers from SAT 1C also trip open, disconnecting all nonessential loads from SAT 1 C to preclude overloading of the transformer.

SATs IC and ID are sized to accommodate the simultaneous starting of all required ESF loads on receipt of an accident signal without the need for load sequencing.

However, ESF loads are sequenced when the associated 4.16 kV ESF bus is supplied from SAT 1C.A description of the Unit 2 offsite power sources is provided in the Bases for Unit 2 LCO 3.8.1, "AC Sources -Operating." (continued)

HATCH UNIT I B 3.8-1 REVISION I HATCH UNIT 1 B 3.8-1 REVISION 1 V1 3 Page 2 of 5 AC Sources -Operating B 3.8.1 BASES BACKGROUND 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 ESF I buses I1E and I1G, respectively.

DG 1 B (the swing DG) is a shared I 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 drywell pressure 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 bus undervoltage or degraded voltage, independent of or coincident with a LOCA signal. The DGs also start and operate in the standby mode without tying to the ESF bus on a LOCA signal alone. Following the trip of offsite power, load shed relays strip nonpermanent loads from the ESF bus. When the DG is tied to the ESF bus, loads are then sequentially connected to its respective ESF bus by the automatic load sequence timing devices. The sequencing logic controls the permissive and starting signals to motor breakers to prevent overloading the DG.In the event of a loss of preferred power, the ESF electrical loads are automatically connected to the DGs in sufficient time to provide for safe reactor shutdown and to mitigate the consequences of a Design Basis 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 safe condition are returned to service (i.e., the loads are energized.)

DGs IA, 1B, and IC have the following ratings: a. 2850 kW- 1000 hours0.0116 days 0.278 hours 0.00165 weeks 3.805e-4 months , and b. 3250 kW- 168 hours0.00194 days 0.0467 hours 2.777778e-4 weeks 6.3924e-5 months .A description of the Unit 2 onsite power sources is provided in the Bases for Unit 2 LCO 3.8.1.(continued)

HATCH UNIT I1 .- RVSO B 3.8-2 REVISION 1 V1 3 Page 3 of 5 AC Sources -Operating B 3.8.1 B 3.8 ELECTRICAL POWER SYSTEMS B 3.8.1 AC Sources -Operating BASES BACKGROUND The Unit 2 Class 1 E AC Electrical Power Distribution System AC sources consist of the offsite power sources (preferred power sources, 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 power system provides independence and redundancy to ensure an available source of power to the Engineered Safety Feature (ESF)systems.The Class 1 E AC distribution system is divided into redundant load groups, so loss of any one group does not prevent the minimum safety functions from being performed.

Each load group has connections to two preferred offsite power supplies and a single DG.Offsite power is supplied to the 230 kV and 500 kV switchyards from the transmission network by eight transmission lines. From the 230 kV switchyards, two electrically and physically separated circuits provide AC power, through startup auxiliary transformers 2C and 2D, to 4.16 kV ESF buses 2E, 2F, and 2G. A detailed description of the offsite power network and circuits to the onsite Class 1 E ESF buses is found 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 power from the offsite transmission network to the onsite Class 1 E ESF bus or buses.Startup auxiliary transformer (SAT) 2D provides the normal source of power to the ESF buses 2E, 2F, and 2G. If any 4.16 kV ESF bus loses power, an automatic transfer from SAT 2D to SAT 2C occurs.At this time, 4.16 kV buses 2A and 2B and supply breakers from SAT 2C also trip open, disconnecting all nonessential loads from SAT 2C to preclude overloading of the transformer.

SATs 2C and 2D are sized to accommodate the simultaneous starting of all required ESF loads on receipt of an accident signal without the need for load sequencing.

However, ESF loads are sequenced when the associated 4.16 kV ESF bus is supplied from SAT 2C.A description of the Unit 1 offsite power sources is provided in the Bases for Unit 1 LCO 3.8.1, "AC Sources -Operating." (continued)

I HATCH UNIT 2 B38IRVSO B3.8-1 REVISION 1 V1 3 Page 4 of 5 AC Sources -Operating B 3.8.1 BASES BACKGROUND 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 ESF buses 2E and 2G, respectively.

DB 1 B (the swing DG) is a shared power source and can supply either Unit I ESF bus IF or Unit 2 ESF bus 2F. A OG starts automatically on a loss of coolant accident (LOCA) signal (i.e., low reactor water level signal or high drywell pressure 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 bus undervoltage or degraded voltage, independent of or coincident with a LOCA signal. The DGs also start and operate in the standby mode without tying to the ESF bus on a LOCA signal alone. Following the trip of offsite power, load shed relays strip nonpermanent loads from the ESF bus. When the DG is tied to the ESF bus, loads are then sequentially connected to its respective ESF bus by the automatic load sequence timing devices. The sequencing logic controls the permissive and starting signals to motor breakers to prevent overloading the DG.In the event of a loss of preferred power, the ESF electrical loads are automatically connected to the DGs in sufficient time to provide for safe reactor shutdown and to mitigate the consequences of a Design Basis 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 safe condition 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, and d. 3500 kW- 30 minutes.DG I1B has the following ratings: a. 2850 kW -1000 hours, and (continued)

HATCH UNIT 2 B382RVSO B 3.8-2 REVISION 1 V1 3 Page 5 of 5 2 3Okv Ma in Power Transformer No.1 1200 MA @ 65oC ODAF No. 1 -A.w,,;rf 105OMVA ,mo T 0.88PF u(,UoTRALI , e T,,u--' +t lUS:S of oad o 410-Vbu es. ms REV 33 9115~4160-V AUXILIARY SOUTHEN SOUITHERN NUCLEAR OPERATING COMPANY ELECTRICAL POWER SYSTEM COMANYu~r EDINI. HATCH NUCLEAR PLANT,o.W,, o, FIGURE 8.3-1 V14 Page 1 of 1 Hatch -Rx Vessel Pressure Instrumentation V1 5 Page 1 of 4 DC Sources -Operating B 3.8.4 BASES BACKGROUND (continued) result in the discharging of the associated battery (and affect the battery cell parameters).

The DC power distribution system is described in more detail in Bases for 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 load continuously for approximately 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months (Ref. 4).Each DC battery subsystem is separately housed in a ventilated room apart from its charger and distribution panels. Each subsystem is located in an area separated physically and electrically from the other subsystems to ensure that a single failure in one subsystem does not cause a failure in a redundant subsystem.

There is no sharing between redundant Class 1 E subsystems such as batteries, battery chargers, or distribution panels.The batteries for DC electrical power subsystems are sized to produce required capacity at 80% of nameplate rating, corresponding to warranted capacity at end of life. The minimum design voltage limit is 105/210 V.Each battery charger of DC electrical power subsystem has ample power output capacity for the steady state operation of connected loads required during normal operation, while at the same time maintaining a fully charged battery. Each battery charger has sufficient capacity to restore the battery from the design minimum charge to its fully charged state within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months while supplying normal steady state loads (Ref. 4).A description of the Unit 2 DC power sources is provided in the Bases for Unit 2 LCO 3.8.4, "DC Sources -Operating." APPLICABLE SAFETY ANALYSES The 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 are OPERABLE.

The DC electrical power system provides normal and emergency DC electrical power for the DGs, emergency auxiliaries, and control and switching during all MODES of operation.

The OPERABILITY of the DC subsystems is consistent with the initial assumptions of the accident analyses and is based upon meeting the design basis of the unit. This includes maintaining DC sources OPERABLE during accident conditions in the event of: (continued)

HATCH UNIT I1 .-3REIIN3 B 3.8-53 REVISION 33 V1 5 Page 2 of 4 HNP-1 -FSAR-8 8.

5.3 DESCRIPTION

wro separate plant batteries are furnished, each with its own static-type battery chargers, circuitI reakers, and bus. One spare battery charger is provided for each of the two batteries for enticing and to back up the two normal power supply chargers.

Plant battery operating voltageI 1s/20 .lach battery with its main dc bus is in a separate room separated by a concrete wall. 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 fire from one room into the other.Batteries (IA and I B) are 120-cell lead-calcium type with a continuous discharge rating of 1410OAh and 1513 Ah, respectively, for 2 h at 77°F to 1.75 V final average cell voltage. These batteries are not tested at the 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months rate.All six 125-V-dc battery chargers are full-wave silicon-controlled rectifier type rated 400 A with an output voltage regulation of + 0.75% from no load to 2% load and + 0.5% from 2% load to full load, 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 bus is 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 to carry 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 its associated 4-kV bus. (See drawing no. H-I13371 .) Each battery has its own SCR type battery charger, circuit breaker, and bus with a spare battery charger for each battery to permit servicing or sparing any charger. Emergency battery operating voltage is 125 V.Control power for each diesel generator, its generator breaker, and the associated 4-kV switchgear bus power feeder circuit breakers is supplied by its respective battery. Diesel battery 1A also supplies control power for 4160 V switchgear bus 1 E and Division I loads on bus 1 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-V switchgear 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 load for an approximate 2-h period without recharging.

These batteries are 60-cell lead-calcium type with a discharge rating of 410 Ah for batteries IA and 1 C and 495 Ah for battery 1 B for 8 h at 77°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 voltage regulation of + 0.75% from no load to 2% load and + 0.5% from 2% load to full load with ac supply variation of + 10% in voltage and + 5% in frequency.

8.5-2 8.5-2 REV 28 9/10 V1 5 Page 3 of 4 2R23-SO03 600 V BUS C)KL m 600 208/1 ,OV 125/255 V BATTERY BATrERY CHARGERS J[)PHASE 2 PRIMARY STRATEGY 600 KW FLEX DG lX86-5503 ESSENTIAL 2R25 CAB A 5036)2R22-SO16 125 V DC BUS DIV I INSTRUMENT BUS 2A 2R25-SO64 600 V MCC C PHASE 2 ALTERNATE STRATEGY 600 KW FLEX DG 1X86-S003 cables connected directly to charger disconnects RHR RHR MOVs RM COOLER Vi 5 Page 4 of 4 I BATTERY ICHARGERS IPHASE 2_______ ____PRIMARY STRATEGY ESSENTIAL CAB B 125/250 V 600 KW FLEX DG 2R2= .S037 BATTERY 2R22-S017 125 V DC BUS DIV II 1X86-S004 I> 600 VMCC D PHASE 2 600 K'W FLEX DG))1X86-S004 cables connected directly to charger disconnects RHR RHR MOVs INSTRUMENT BUS 2B RM COOLER 2R25-S065 SMNH-1 3-021 Attachment L

f..lIr~idItinn fnr F:-W~ll SHEET L-6 v16 Page 1 of 3 CALC NO. SNCO24-CALC-007 HNP DETERMINATION OF EMERGENCY ACTION LEVEL REV. 0 0I E N E R C 0 N FOR INITIATING CONDITION E-PAGE NO. Page 6of 8 5.0 Design Inputs 1. The contact dose rates from the HI-STORM 100 and HI-TRAC 125 cask system technical specification

[2, Table 6.2-2] are provided below in Table 5-1. These source 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-STORM 100 and HI-TRAC 125 LaatonNumber of Technical SpecIfication Measurements Limit (mrem/hr)H I-TRAC 125 _________Side -Mid -height 4 224.9 Top 1 4 J52.8 HI-STORM 100 _________Side -60 inches below mid-height 4 38.9 Side -Mid -height 4 39.7 Side -60 inches above mid-height 4 15.6 Top -Center of lid 1 6.0 Top -Radially centered 4 8.4 Inlet duct 4 72.0 Outlet duct 4 18.6 Attachment L FNFRCPr4N (CAleIlItinn fnr F-Hll SMNH-1 3-021 SHEET L-7 v16 Page 2 of 3 CALC NO. SNCO24-CALC-007 HNP DETERMINATION OF -___________

EMERGENCY ACTION LEVEL REV. 0 I0 E N E R C 0 N FOR INITIATING CONDITION E-" HU1 PAGE NO. Page 7oft8 6.0 Methodology The "on-contact"'

dose rates from the technical specification for the HI STORM-i100 and HI-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 3-021 Attachment L ENERCON Calculation for E-HU1 SHEET L-8 v16 Page 3 of 3 CALC NO. SNCO24-CALC-007 HNP DETERMINATION OF ___________

EMERGENCY ACTION LEVEL REV. 0 0O E N E R C 0 N FOR INITIATING CONDITION E- -_________HU1 PAGE NO. Page 8of 8 7.0 Calculations The dose rates in Table 5-1 are multiplied by 2 in order to calculate the EAL dose rate limits. 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 EAL LoainLimit Factor (mrem/hr) (mrem/hr)________________________ (mrem/hr)

____________________

_______ ______ ______ ______HI-TRAC 125 _ _ _ _ _ _ _ _ _ _ _ _ _ _Side -Mid -height 1 224.9 f 2 449.8 j 450 Top J 52.8 j 2 105.6 110 HI-STORM 100 ___________

Side -60 inches below mid-height 38.9 2 77.8 80 Side -Mid -height 39.7 2 79.4 80 Side -60 inches above mid-height 15.6 2 31.2 30 Top -Center of lid 6.0 2 12 10 Top -Radially centered 8.4 2 16.8 20 Inlet duct 72.0 2 144 140 Outlet duct 18.6 2 37.2 40 8.0 Computer Software Microsoft WORD 2013 is used in this calculation for basic multiplication.

V17 Page 1 of 2 1.0 IDENTIFICATION:

ALARM PANEL 601-3______

DRY WELL HIGH PRESSURE INfITATION 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 O Primary 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 Drywell m6.2 Excessive N 2 inerting 6.3 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 Pressure 34AR-601 -305-1 VER. 5.2 MGR-0048 Ver. 5.0 V17 Page 2 of 2 1.0 IDENTIFICATION:

ALARM PANEL 601-3 ______DRY WELL HIGH PRESSURE INITIATION DEVICE: SETPOINT: 2E11-PIS-N694N1B/C/D 1.85 PSIG 2.0 CONDITON:

13.0 CLASSIFICATION:

EMERGENCY A high pressure condition exists In the Drywell. 4.0 LOCATION: 2H11 -P601 Panel 601-3 5.0 OPERATOR ACTIONS: 5.1 Confirm a high Drywell pressure condition exists on 2T48-R607A/2T48-R607B Narrow Range 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 ANDD enter 31 EO-EOP-010-2, RC RPV Control (Non-ATWS).

[J 5.3 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 Drywell 6.2 Excessive N 2 inerting 6.3 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 3.3.5.1 System Eli Elementary Diagrams 8.2 TS 3.5.1 7.2 57CP-CAL-102-2, Analog Master/Slave Trip Unit Cal 34AR-601 -302-2 Ver. 3.2 C MGR-0048 Ver. 5.0AGMR7-10 AG-MGR-75-1101 V6 Page 1 of 6 SOUTHERN NUCLEAR DOCUMENT TYPE: PLANT E.I. HATCH ANNUNCIATOR RESPONSE PROCEDURE (ARP) AGE 1 OF 2 DOCUMENT 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/A ANNUNCIATOR RESPONSE PROCEDURES FOR 1H11-P654 -ALARM PANEL 1 ARP NO.VER NO.6.1 ARP NO.VER NO.ARP NO.VER. NO.SPARE 34AR-654-001

-1 34AR-654-031-1 4.0 34AR-654-061 34AR-654-002-1 1.0 34AR-654-032 SPARE 34AR-654-062-1 3.0 34AR-654-003-1 4.1 34AR-654-033-1 SPARE 34AR-654-063-1 2.0 34AR-654-004-1 6.1 34AR-654-034-1 2.0 34AR-654-064-1 3.0 34AR-654-005-1 3.0 34AR-654-035-1 2.0 34AR-654-065-1 7.3 34AR-654-006-1 7.1 34AR-654-036-1 1.0 34AR-654-066-1 2.0 34AR-654-007-1 12.0 34AR-654-037-1 4.0 34AR-654-067-1 8.0 34AR-654-008-1 4.0 34AR-654-038-1 4.0 34AR-654-068-1 8.0 34AR-654-009-1 5.2 34AR-654-039-1 2.0 34AR-654-069-1 2.0 34AR-654-0 10-1 2.1 34AR-654-040-1 2.0 34AR-654-070-1 3.1 34AR-654-01 1 SPARE 34AR-654-041-1 2.1 34AR-654-071-1 2.0 34AR-654-012-1 0.0 34AR-654-042-1 1.0 34AR-654-072-1 2.0 34AR-654-01 3-1 2.0 34AR-654-043-1 2.0 34AR-654-073-1 2.0 34AR-654-014-1 3.0 34AR-654-044-1 3.0 34AR-654-074-1 3.0 34AR-654-015 SPARE 34AR-654-045 SPARE 34AR-654-075-1 SPARE 34AR-654-0 16-1 4.0 34AR-654-046-1 3.0 34AR-654-076-1 3.0 34AR-654-017 SPARE 34AR-654-047-1 1.0 34AR-654-077-1 1.0 34AR-654-01 8-1 3.1 34AR-654-048-1 3.0 34AR-654-078-1 1.1 34AR-654-01 9-1 2.0 34AR-654-049-1 3.0 34AR-654-079-1 2.0 34AR-654-020-1 2.0 34AR-654-050-1 4.1 34AR-654-080-1 6.2 34AR-654-02 1-1 5.4 34AR-654-05 1-1 3.0 34AR-654-08 1-1 SPARE 34AR-654-022-1 11.1 34AR-654-052

-SPARE 34AR-654-082 SPARE 34AR-654-023-1 4.0 34AR-654-053

.SPARE 34AR-654-083 SPARE 34AR-654-024-1 2.0 34AR-654-054-1 3.0 34AR-654-084-1 2.0 34AR-654-025 SPARE 34AR-654-055-1 3.0 34AR-654-085-1 3.0 34AR-654-026 SPARE 34AR-654-056-1 1.1 34AR-654-086-1 2.0 34AR-654-027 SPARE 34AR-654-057-1 1.0 34AR-654-087-1 2.0 34AR-654-028-1 1.0 34AR-654-058-1 2.0 34AR-654-088-1 2.0 34AR-654-029-1 2.0 34AR-654-059-1 3.0 34AR-654-089-1 3.0 34AR-654-030 SPARE 34AR-654-060-1 0.0 34AR-654-090-1 SPARE NOTE. 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. j I Level Of Use ARPs CONTINUOUS ALL REFERENCE None INFO None NMP-AP-002 V6 Page 2 of 6 PAGE 2 OF2 SOUTHERN NUCLEAR PLANT E.I. HATCH DOCUMENT TITLE: DOCUMENT NUMBER: ARP'S FOR CONTROL PANEL 1HI11-P654,I 34AR-654-901-1 ALARM PANEL 1 VERSION NO: 23.0 UNIT 1 1 H11-P654 -1 (Left)654-00 1 654-002 654-003 654-004 654-005 654-006 654-007 654-008 654-016 654-017 654-018 654-019 654-020 654-021 654-022 654-023 654-03 1 654-032 654-033 654-034 654-035 654-036 654-037 654-038 654-046 654-047 654-048 654-049 654-050 654-05 1 654-052 654-053 654-06 1 654-062 654-063 654-064 654-065 654-066 654-067 654-068 654-076 654-077 654-078 654-079 654-080 654-08 1 654-082 654-083 UNIT I 1H11-P654

-1 (Right)654-009 654-010 654-011 654-012 654-013 654-014 654-015 654-024 654-025 654-026 654-027 654-028 654-029 654-030 654-039 654-040 654-04 1 654-042 654-043 654-044 654-045 654-054 654-055 654-056 654-057 654-058 654-059 654-060 654-069 654-070 654-07 1 654-072 654-073 654-074 654-075 654-084 654-085 654-086 654-087 654-088 654-089 654-090 NMP-AP-002 V6 Page 3 of 6 1.0 IDENTIFICATION:

ALARM PANEL 654 SPENT FUEL STORAGE POOL LEVEL LOW DEVICE: 1G41-N362 Level Sensor 1G41-N372 Remote Electronics SETPOINT: 225 ft. 9 in.(21' -7" above the top of the fuel assemblies seated in the Fuel Pool)5.0 OPERATOR ACTIONS:* 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.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 damage the DHR pump.5.1 Enter 34AB-G41-002-1, Decreasing Rx Well/Fuel Pool Water Level. Li 5.2 IF Fuel Pool gates are installed:

5.2.1 Raise water level by regulating 1G41-F041, Spent Fuel Pool Make-up water from CST Valve, located at 185RBR07, panel 1H21-P155.

LI 5.2.2 Confirm Fuel Pool Cooling filter effluent is returning to fuel pool per 34SO-G41-003-1, Fuel Pool Cooling and Cleanup System. LI 5.3 IF Fuel Pool gates are NOT installed, request Maintenance to install gates. LI 5.4 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. LI 6.0 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 3.7.8 7.2 H-17074, Fuel Pool Cooling System G41 Elem Diag 34AR-654-022-I VER. 11.1 MGR-0048 Ver. 5.0NMPA02 NMP-AP-002 V6 Page 4 of 6 DOCUMNT TTLE:DOCUMENT NUMBER: VERSION NO: ARP'SALRFOR CONTROLpAE PANEL 2H11-P654, 34AR-654-901-2 23.4 EXPI RATION APPROVALS:

EFFECTIVE DATE: DEPARTMENT MANAGER DAK for KDL DATE 09-07-12 DATE: 2-27-15 N/A SSM /PM N/A DATE N/A _____ANNUNCIATOR RESPONSE PROCEDURES FOR 2H11-P654

-ALARM PANEL 1 ARP 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 SPARE 34AR-654-002-2 2.1 34AR-654-032-2 2.1 34AR-654-062-2 3.1 34AR-654-003-2 3.1 34AR-654-033 SPARE 34AR-654-063 SPARE 34AR-654-004-2 2.2 34AR-654-034-2 2.3 34AR-654-064-2 2.3 34AR-654-005-2 2.0 34AR-654-035-2 2.1 34AR-654-065-2 4.2 34AR-654-006-2 8.0 34AR-654-036-2 3.4 34AR-654-066-2 5.1 34AR-654-007-2 11.0 34AR-654-037-2 2.0 34AR-654-067-2 6.0 34AR-654-008-2 3.1 34AR-654-038-2 4.1 34AR-654-068-2 6.0 34AR-654-009 SPARE 34AR-654-039 SPARE 34AR-654-069 SPARE 34AR-654-01 0 SPARE 34AR-654-040-2 3.0 34AR-654-070-2 4.2 34AR-654-01 1 SPARE 34AR-654-041-2 3.1 34AR-654-071-2 4.1 34AR-654-012-2 3.2 34AR-654-042-2 2.1 34AR-654-072-2 4.1 34AR-654-013-2 2.2 34AR-654-043-2 2.1 34AR-654-073-2 4.1 34AR-654-014-2 1.2 34AR-654-044-2 2.1 34AR-654-074-2 5.4 34AR-654-015-2 3.2 34AR-654-045 SPARE 34AR-654-075 SPARE 34AR-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 2.2 34AR-654-020-2 3.2 34AR-654-050 SPARE 34AR-654-080-2 7.5 34AR-654-021-2 7.1 34AR-654-051 SPARE 34AR-654-081 SPARE 34AR-654-022-2 11.1 34AR-654-052 SPARE 34AR-654-082 SPARE 34AR-654-023-2 6.1 34AR-654-053 SPARE 34AR-654-083 SPARE 34AR-654-024 SPARE 34AR-654-054 SPARE 34AR-654-084 SPARE 34AR-654-025 SPARE 34AR-654-055-2 3.0 34AR-654-085-2 4.2 34AR-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 4.1 34AR-654-029 SPARE 34AR-654-059-2 2.2 34AR-654-089 SPARE 34AR-654-030 SPARE 34AR-654-060 SPARE 34AR-654-090 SPARE I I I INOTQE: I Approval signature on this page constitutes approval for all procedures listed above at the Iversion indicated.

Tab numbers in the back correspond to procedure sequence number.I lit rm Level Of Use ARPs CONTINUOUS ALL REFERENCE None INFO None V6 Page 5 of 6 2H11-P654-1 (Left)654-001 654-002 654-003 654-004 654-005 654-006 654-007 654-008 654-016 654-017 654-018 654-019 654-020 654-021 654-022 654-023 654-031 654-032 654-033 654-034 654-035 654-036 654-037 654-038 654-046 654-047 654-048 654-049 654-050 654-051 654-052 654-053 654-061 654-062 654-063 654-064 654-065 654-066 654-067 654-068 654-076 654-077 654-078 654-079 654-080 654-081 654-082 654-083 UNIT 2 2H11-P654

-1 (Right)654-009 654-010 654-011 654-012 654-013 654-014 654-015 654-024 654-025 654-026 654-027 654-028 654-029 654-030 654-039 654-040 654-041 654-042 654-043 654-044 654-045 654-054 654-055 654-056 654-057 654-058 654-059 654-060 654-069 654-070 654-071 654-072 654-073 654-074 654-075 654-084 654-085 654-086 654-087 654-088 654-089 654-090 V6 Page 6 of 6 1.0 IDENTIFICATION:

ALARM PANEL 654 ________" --t---SPENT FUEL STORAGE POOL LEVEL LOW DEVICE: SETPOINT: 2G41-N372 Remote Electronics 226' 2.5" (approx. 22' 0.5" above the top of 2G41-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. Li 5.2 IF Fuel Pool gates are installed: