ML16071A158

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

ML16071A158
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: ML16071A158 (52)

Text

V9Page 1 of 5Southern Nuclear Operating CompanyI rlll Plant: VEGP TteNE990Re6EACacliosI X6CNA1 5Unit: 1&2 TteNE9-0Re6ELCacliosSHEET 10Volume fraction above operating deck = 0.771

Reference:

Table 6.5.2-1, VEGP FSAR Revision 19 (February 2014)8. Containment liner: 1/4"4 carbon steel

Reference:

VEGP FSAR sections 1.2.5, 6.2.7.2,

& 6.5.3.1 and drawings 1X2D01A001

&2X2D01A001 Reactor Coolant System Parameters

9. Reactor Pressure Vessel & RCS Piping Dimensions Parameter Value Reference RPV Inside Diameter 173" VEGP FSAR Table 5.3.3-1Hot Leg centerline elevation 187'-0" AX4DR023, 1X4DL4A017-1,

&(76% RVLIS) 2X4DL4A01 7-1Cold Leg centerline elevation Hot Leg Nozzle Bottom 185'-91/2' AX4DR023Top of Active Fuel 181'-10" AX4DR023(63% RVLIS)Cold Leg Pipe ID 27%" 1X4DL4A017-1

& 2X4DL4A017-1 Hot Leg Pipe ID 29" 1X4DL4A017-1

& 2X4DL4A017-1 RCS Coolant Parameters Parameter Value Reference Full Power Tavg 588.4 °F Table 2-1, page 2-3, WCAP-16736-P VEGP FSAR Table 15.0.3-3RCS operating pressure 2250 psiaFull power coolant mass 2.53E+08 g Page 3 of LTR-CRA-06-1 79 attachedto WEC-SNC letter GP-1 8006 andTable 7.8-3 of WCAP-16736-P 10.11. Fuel Assembly outside dimensions

= 8.424" x 8.424"

Reference:

1 X6AN09-1 0000-2 & 2X6AN09-1 0000-012. Core effective diameter

= 132.7 inches x 1 foot/12 inches = 11.06 ft

Reference:

Table 5-1, page 5-4, 1/2X6AA10-00095 Source Terms V9Page 2 of 5Southern Nuclear Operating CompanySOI AmiE Plant: VEGP ITitle: NEI 99-01 Rev 6 EAL Calculations X6CNA15~I Unit: 1&2 SHEET 37CA1: Loss of RPV inventory.

Operating Mode Applicability:

Emergency Action Levels:Cold Shutdown, Refueling 1 OR21. Loss of RPV inventory as indicated be level less than elevation 185'-10" (73% on FullRange RVLIS).The RPV water level elevations corresponding to the RCS loop piping bottom IDsare found as follows:Dimension IElevation Loop Centerline Elevation 1 87'-00"Cold LegInside Diameter 27.5"1/2AxID 13.75"Bottom ID = Centerline

-(1/2Ax ID) 185'-1 0.25"Hot LegInside Diameter 29.0"1/2Ax ID 14.5"Bottom ID = Centerline

-(A x ID) 185'-9.5" The dimensions and elevations are taken from Design Input #9. The RPV waterlevel elevation corresponding to the Bottom ID of the RCS piping is ~185'1O".

Because the core barrel is a right circular

cylinder, the RVLIS indication corresponding to the above RPV water level can be determined by linearlyinterpolating between the TOAF (EL 181'-10" or 63% RVLIS) and the CL and HLcenterline elevation (EL 187"0O" or 76% RVLIS):

V9Page 3 of 5Southern Nuclear Operating CornpanyouIu A i= Plant: VEGP ITteNE990Re6EACacliosX6CNA1 5 II ULOMPA Unit: 1&2 Til:NI9-Rv6ELCluaion SHEET 38VEGP RVLIS Indication vs. RPV Water Level Elevation

} .i181 182 F 18 8 8 8 88RPV Water Level Elevation (feet)The RPV water level elevation corresponding to the Bottom ID is 185'-10" or~73% on Full Range RVLIS.2. a. RPV level cannot be monitored for 15 minutes or longerANDb. UNPLANNED increase in Containment sump, Reactor Coolant Drain Tank (RCDT)or Waste Holdup Tank (WHT) levels due to a loss of RPV inventory.

V9Page 4 of 5Southern Nuclear Operating Company4xnlmM. Plant: VEGP ITitle: NEI 99-01 Rev 6 EAL Calculations X 6CNA15 ISO mUTH t Unit: 1&2 S HEET 39CSI: Loss of RPV inventory affecting core decay heat removal capability.

Operating Mode Applicability:

Emergency Action Levels:Cold Shutdown, Refueling 1 OR20OR31. a. CONTAINMENT CLOSURE not established ANDb. RPV water level less than 1 85'-4" [6" below Bottom ID of loop] (72% on Full RangeRVLIS).The RPV water level elevations corresponding to 6" below the cold leg (CL) andhot leg (HL) bottom IDs are found as follows:Dimension Elevation Loop Centerline Elevation 1 87'-00"Cold LegInside Diameter 27.5"% x ID 13.75"Bottom ID = Centerline

-(1/2 x ID) 185'-10.25" 6" Below CL Bottom ID 1 85'-4.25" Hot LegInside Diameter 29.0"% xID 14.5"Bottom ID = Centerline

-(1/2 x ID) 185'-9.5" 6" Below HL Bottom ID 1 85'-3.5"The dimensions and elevations are taken from Design Input #9. The elevation corresponding to 6" below the Bottom ID of the RCS piping is ~185'4".Because the core barrel is a right circular

cylinder, the RVLIS indication corresponding to the above RPV water level can be determined by linearlyinterpolating between the TOAF (EL 181 '-10" or 63% RVLIS) and the CL and HLcenterline elevation (EL 187"-0" or 76% RVLIS):

V9Page 5 of 5Southern Nuclear Operating CompanyPlnt: VEGP Title: NEI 99-01 Rev 6 EAL CalculationsX6N1 mpw Unit: 1&2 I SHEET 40VEGP RVLIS Indication vs. RPV Water Level Elevation 621 "B~owRCS-

__ ___.... .... ..... i .jPiping BottomU)181 182 183 184 185 188 187 188RPV Water Level Elevation (feet)The RPV water level elevation corresponding to 6" below the Bottom ID is185'-4" or -72% on Full Range RVLIS.2. a. CONTAINMENT CLOSURE established ANDb. RPV level less than 181'-1 0" ITOAF] (63% on Full Range RVLIS).3. a. RPV level cannot be monitored for 30 minutes or longerANDb. Core uncovery is indicated by ANY of the following:

RE-005 O..R 006 > 40 REM/hrErratic Source Range monitor indication UNPLANNED increase in Containment Sump, Reactor CoolantDrain Tank (RCDT) or Waste Holdup Tanks (WHT) levels ofsufficient magnitude to indicate core uncovery Vl 0VEGP-FSAR-1 1Pae1o511.2.1.3 Eqluipment DesignThe LWPS equipment design parameters are provided in table 11.2.1-2.

The seismic design classification and safety classification for the LWPS components andstructures are listed in table 3.2.2-1.

Safety class designations are also indicated on the LWPSpiping and instrumentation

diagram, drawings 1X4DB 124, 1X4DB 125, 1X4DB 126, 1X4D B127,AX4DB1 24-2, AX4DB 124-3, AX4DB1 24-4, and AX4DB 124-5.11.2.1.4 Reference

1. U.S. Nuclear Regulatory Commission, "Calculation of Releases from Pressurized WaterReactors,"

NUREG-0017, April 1976.11.2.2 SYSTEM DESCRIPTIONS The liquid waste processing system (LWPS) collects and processes potentially radioactive wastes for recycling or release to the environment.

Provisions are made to sample and analyzefluids before discharge.

Based on the laboratory

analysis, these wastes are either retained forfurther processing or released under controlled conditions through the cooling water system,which dilutes the discharge flow. A permanent record of liquid releases is provided by analysesof known volumes of effluent.

The radioactive liquid discharged from the reactor coolant system (RCS) is processed by theradwaste processing facility systems and may be discharged or recycled.

The LWPS is arranged to recycle reactor grade water if desired.

This is implemented by thesegqregqation of equipment drains and waste streams to prevent intermixingq of liquid wastes.The LWPS can be divided into the following subsystems:

A. Reactor Coolant Drain Tank (RCDT) Subsystem This portion of the LWPS collects nonaerated, reactor grade effluent fromsources inside the containment.

B. Drain Channel AThis portion of the LWPS collects

aerated, reactor grade effluent that can berecycled.

C. Drain Channel BThis portion of the LWPS processes all effluent that is not suitable for recycling.

D. Radwaste Processing Facility Demineralizers The radwaste processing facility demineralizer systems consist of portabledemineralizers installed in subterranean enclosures inside the radwasteprocessing facility.

The radwaste processing facility is described in paragraph 11.4.2.4.

The radwaste processing facility demineralizers can be aligned toprocess any of the three waste drain streams.E. The radwaste processing facility filtration system consists of a portable, vendorsupplied system located within a shielded area inside the radwaste processing facility.

The filtration system associated tanks, pumps, accumulator, piping,valves, and controls located within a shielded area inside the radwaste11.2-411.2-4REV 13 4106 Vl0aPage 2 of 5VEGP-FSAR-1 1processing facility.

The peripheral equipment is located adjacent to the filterassembly.

The filter system can be aligned to process any of the three wastedrain streams.

Details of this equipment are shown on drawing AX4DB1 24-1.In addition, the LWPS provides capability for handling and storage of spent ion exchangeresins.The LWPS does not include provisions for processing secondary system wastes. Secondary system effluent is handled by the steam generator blowdown processing system (SGBPS),

asdescribed in subsection 10.4.8, and by the turbine building drain system. Estimated releasesfrom these systems are discussed in subsection 11.2.3. The LWPS design, which segregates primary and secondary wastes, minimizes the amount of water that must be processed bydischarging low activity wastes directly, where permissible, with no treatment.

Instrumentation and controls necessary for the operation of the LWPS are located on a controlboard in the auxiliary building.

Any alarm on this control board (except for the waste processing holdup control panel) is relayed to the main control board in the control room.The LWPS piping and instrumentation diagrams are shown in drawings

1X4DB124, 1X4DB125,

1X4DB126, 1X4DB127, AX4DB124-1, AX4DB124-2, AX4DB124-3, AX4DB124-4, andAX4DB1 24-5 and process flow diagram for the LWPS is shown on figure 11.2.2-1.

Table11.2.1-1 lists the assumptions regarding flows and activity levels that were used in preparation of table 11.2.1-3, which gives nuclide concentrations for key locations within the LWPS asshown on figure 11.2.2-1.

The process flow data is calculated using the data in table 11.2.1-1, the flow paths indicated on figure 11.2.2-1, realistic primary coolant activity levels from section11.1, and decontamination factors as given in reference 1 of subsection 11.2.1.11.2.2.1 Reactor Coolant Drain Tank Subsystem IRecyclable reactor grade effluents enter this subsystem from valve leakoffs, reactor coolantIpump No. 2 seal leakoffs, reactor vessel flange leakoff, and other deaerated, tritiated waterIsources inside the containment.

Connections are provided for draining the RCS loops and thesafety injection system (SIS) accumulators and for cooling the pressurizer relief tank. Inaddition, refueling canal drains can be routed to the refueling water storage tank using theRCDT pumps.The RCDT contents are continuously recirculated through the RCDT heat exchanger tomaintain the desired temperature.

Level is prevented from varying significantly by a controlvalve which automatically opens a path from the recirculation line to the BRS when normal tanklevel is exceeded.

The RCDT is also connected to the gaseous waste processing system(GWPS) vent header. Hydrogen gas bottles connected to the RCDT ensure a hydrogenblanket.

Maintaining a constant level minimizes the amount of gas sent to the GWPS andminimizes the amount of hydrogen used. Provisions for sampling the gas are provided.

Details of the RCDT subsystem are shown on drawing 1X4DB127.

A separate RCDTsubsystem is provided for each of the two units.11.2.2.2 Drain Channel A Subsystem

Aereated, tritiated liquid enters drain channel A through lines connected to the waste holduptank. Sources of this aerated liquid are as follows:A. Accumulator drainage (via RCDT pump suction).

11.2-511.2-5REV 13 4/06 V10oPage 3 of 5VEGP-FSAR-11I B. Sample room sink drains (excess primary sample volume only).C. Ion exchanger, filter, pump, and other equipment drains.The containment sump or auxiliary building sump may be directed to the waste holdup tank orthe floor drain tank for processing as necessary.

The collected aerated drainage is pumped or flows to the waste holdup tank prior to processing through the radwaste processing facility filtration system and/or the radwaste processing facilitydemineralizers before reuse or discharge.

Details of this equipment are shown on drawingsAX4DB1 24-2, AX4DB1 24-3, AX4DB1 24-4, and AX4DB1 24-5.The basic composition of the liquid collected in the waste holdup tank is boric acid and waterwith some radioactivity.

A separate drain channel A subsystem is provided for each of the two units. Details are shownon drawings 1X4DB124 and 1X4DB127.

Table 11.2.1-1 lists the estimated flows entering thewaste holdup tank.11.2.2.3 Drain Channel B Subsystem Drain channel B is provided to collect and process nonreactor grade liquid wastes. Theseinclude:* Wastes from floor drains.* Equipment drains containing nonreactor grade water.* Laundry and hot shower drains.* Other nonreactor grade sources.Drain channel B is comprised of three drain subchannels, each associated with one of thefollowing tanks.A. Laundry and Hot Shower TankThe laundry and hot shower tank is provided to collect and process wasteeffluents from the plant laundry and personnel decontamination showers andhand sinks.Laundry and hot shower drains normally need no treatment for removal ofradioactivity.

This water is transferred to a waste monitor tank through thelaundry and hot shower tank filter for eventual discharge.

If sample analysisindicates that decontamination is necessary, the water can be directed throughthe Unit 1 or Unit 2 waste monitor tank demineralizer or the radwaste processing facility for cleanup.The laundry and hot shower tank and filter are shared by the two units. Detailsof this portion of the LWPS are shown on drawing 1X4DB126.

Table 11.2.1-1lists estimated flows entering the laundry and hot shower tank.B. Floor Drain TankWater may enter the floor drain tank from system leaks inside the containment through the containment sump, from system leaks in the auxiliary buildingthrough auxiliary building sumps and the floor drains, and floor drains in the11.2-611.2-6REV 13 4/06 v10oPage 4 of 5VEGP-FSAR-1 1radwaste facilities.

Sources of water to the containment sump and auxiliary building sumps and floor drains are the following:

1. Fan cooler leaks.2. Secondary side steam and feedwater leaks.3. Primary side process leaks.4. Decontamination water.The containment sump or auxiliary building sumps may be directed to the wasteholdup tank.Another source of water to the floor drain tank is the chemical laboratory drains.Excess nonreactor grade samples that are not chemically contaminated andlaboratory equipment rinse water are drained to the floor drain tank.The contents of the floor drain tank are processed through the radwasteprocessing facility demineralizers and/or the radwaste processing facility filtration system and then pumped to a waste monitor tank for ultimate discharge.

If the activity in the floor drain tank liquid is such that the discharge limits cannotbe met without cleanup, the liquid can be processed by the waste monitor tankdemineralizer, the radwaste processing facility demineralizers, or the radwasteprocessing facility filtration system.A separate floor drain tank and associated equipment are provided for each ofthe two units. Details of this portion of the LWPS are shown on drawing1X4DB126.

Table 11.2.1-1 lists the estimated flows entering the floor drain tank.C. Chemical Drain TankLaboratory samples which contain reagent chemicals (and possibly tritiated liquid) are discarded through a sample room sink which drains to the chemicaldrain tank. Chemical drains requiring radwaste processing are sent to the solidwaste management system or may be processed through the radwasteprocessing facility demineralizers and/or the radwaste processing facility filtration system.The chemical drain tank and associated equipment are shared by Units 1 and 2.Details of this portion of the LWPS are shown on drawing 1X4DB125.

Table11.2.1-1 lists the estimated flow directed to the chemical drain tank.Any liquids released to the environment by the LWPS are first directed to a waste monitor tank.Before releasing the contents of a waste monitor tank, a sample is taken for analysis.

Thefindings are logged, and, if the activity level is within acceptable limits, the tank contents arereleased to the discharge canal. The discharge valve is interlocked with a process radiation monitor and closes automatically when the radioactivity concentration in the liquid discharge exceeds a preset limit. The radiation element is located upstream of the discharge valve at adistance sufficient to close the valve before passing the fluid that activated the detector tripsignal. The isolation valve also blocks flow if sufficient dilution water is not available.

Theradiation monitor is described in section 11.5. A permanent record of the radioactive releasesis provided by a sample analysis of the known volumes of waste effluent released.

Liquidwaste discharge flow and volume are also recorded.

If the monitor tank contents are not acceptable for discharge, the fluid can be held for a time toallow activity to decay to acceptable levels, or it can be further processed by the waste monitor11.2-711.2-7REV 13 4/06 V10Page 5 of 5VEGP-FSAR-11I H. Waste Monitor Tank PumpsTwo pumps are provided for each unit. One pump is used for each monitor tankto discharge water from the LWPS or for recycling if further processing isrequired.

The pump may also be used for circulating the water in the waste monitor tank toobtain uniform tank contents, and therefore a representative sample, beforedischarge.

These pumps can be throttled to achieve the desired discharge rate.I. Auxiliary Waste Monitor Tank PumpsTwo pumps are provided.

They are installed in Unit 2 but serve both units. Onepump is used for each auxiliary waste monitor tank to discharge water fromLWPS or for recycling if further processing is required.

A mixer may be used forcirculating the water in the auxiliary waste monitor tank to obtain uniform tankcontents, thereby assuring a representative sample is acquired prior to discharge of the tank contents.

The pumps can be throttled to achieve the desireddischarge rate.11.2.2.6.2 TanksA. Reactor Coolant Drain TankOne tank is provided for each unit. The purpose of the RCDT is to collectleakoff-type drains inside the containment at a central collection point for furtherdisposition through a single penetration via the RCDT pumps. The tank providessurge volume and net positive suction head (NPSH) to the pumps.Only water which can be directed to the boron recycle holdup tanks enters theRCDT. The water is compatible with reactor coolant and does not containdissolved air during normal plant operation, by engineering design.A constant level is maintained in the tank to minimize the amount of gas sent tothe GWPS and also to minimize the amount of hydrogen cover gas required.

The level is maintained by one continuously running pump and by a control valvein the discharge line. This valve operates on a signal from a level controller tolimit the flow out of the system. The remainder of the flow is recirculated to thetank.Continuous flow is maintained through the heat exchanger in order to preventloss of pump NPSH resulting from a sudden inflow of hot liquid into the RCDT.B. Waste Holdup TankOne atmospheric pressure tank is provided for each unit to collect:1. Equipment drains.2. Valve and pump seal leakoffs (outside the containment).

3. Boron recycle holdup tank overflows.

4. Other water from tritiated, aerated sources.The tank size is adequate to accommodate 11 days of expected influent duringnormal operation.

C. Waste Evaporator Condensate Tank11.2-1111.2-11REV 13 4/06 ViiPage 1 of 3Southern Nuclear Operating CompanysmrllllM~LPlant:

VEGP Title: NEI 99-01 Rev 6 EAL Calculations I 6CNA15¢m Unit: 1&2 SHEET 42UNPLANNED increase in Containment Sump, Reactor Coolant Drain Tank(RCDT) or Waste Holdup Tanks (WHT) level of sufficient magnitude toindicate core uncovery.

ANDc. ANY indication from the Containment Challenge Table Cl (above).Containment Operating Deck High Range (RE-O005 or RE-O006):

This radiation monitor reading corresponds to the reflected dose rate from the irradiated fuel inthe RPV with an RPV water level of less than TOAF (18 1'-10" or 63% on Full RangeRVLIS). It is calculated in Attachment E3 of this calculation.

Erratic Source Range Monitor Indication Basis: NEI 99-01 R6, page 74.Explosive mixture inside containment

> 6% by volume hydrogen:

Sheet 23 of VEGP SAMG calculation X6CNA1 1 established the 6% by volume hydrogen limit.Pressure

> 14 psig WITH CONTAINMENT CLOSURE established:

NMP-EP-1 10-GL03 (pages 88, 90, 922, & 94) defines CONTAINMENT CLOSURE per"Operating Procedure 142 10-1/2, Containment Building Penetrations Verification

-Refueling."

Section 6.0 of 142 10-1/2 lists the acceptance criteria for CONTAINMENT

CLOSURE, amongthem the requirement that >23' of water (EL 21 7'-0") is maintained above the RPV flange. Thiscorresponds to the SEP water low level (Design Input #4).During Refueling Operations, fuel is moved between the Containment and the Fuel HandlingBuilding via the Fuel Transfer Tube. The Fuel Transfer Tube will maintain an air-to-air barrierduring refueling operations if either the gate valve is closed or the water level in the refueling cavity is high enough to provide an air-to-air barrier.

VllPage 2 of 3Southern Nuclear Operating CornpanyAOm~I 4 Plant: VEGP Title: NEI 99-01 Rev 6 EAL Calculations I X6CNA1 5I MV Unit: 1&2 I SHEET 53The results of the Loss of Clad FP Barrier setpoint calculations in Attachments H-3 and 13 aresummarized below. Given the system accuracy

-a factor of two over the operating range -thethreshold is rounded off to two significant figures.Unit Calculated Threshold Rounded-Off Threshold (R EM/h r) (m RE M/h r)VEGP 1 1.31E+04 1.3E+07VEGP 2 1.49E+04 1.5E+07Containment Barrier Potential Loss Threshold 4.BContainment Hydrogen concentration greater than 6%.The existence of an explosive mixture means, at a minimum, that the containment atmospheric hydrogen concentration is sufficient to support a hydrogen burn (i.e., at the lower deflagration limit). A hydrogen burn will raise containment pressure and could result in collateral equipment damage leading to a loss of containment integrity.

It therefore represents a potential loss of thecontainment barrier.Sheet 23 of VEGP SAMG calculation X6CNA II established the 6% by volume hydrogen limit.

VllPage 3 of 3Desis,,C.lulation

-Nuclear Southern Conmpay Services Aim~Prjc:Vogtl lei Gener~atig Plat Ci.No. X6CNAn 1 5SSubJectflitle:

Severe Accident Management Guideline (SAMG) Calculations Sheet 23 of 167CA-3 HYDROGEN FLAMMABILITY IN CONTAINMENT Determined Values: See attached graphs.Guidelines:

SAG-2, 3, 7, SCG-3

References:

1. MUHP-23 10, WOG Severe Accident Management Guidance (Background Document) and MUHP-23 15 I //sWOO Severe Accident Management Guidance.,

Rev. 12. EpRI TR-101 869, Severe Acodient Managemnent Guidance Technical Basis Report, Volume 2: The Physicsof Accident Progression

3. FSAR:a. Secion 6.2.1.5.2

c. Figure 6.2.1-1 INb. Table 6.2.5-6 d. Figur 6.2.1-44. Technical Specifications:

a. Section 3.6.1.4b. Section 3.6.1.5 1/5. Memo from Roger Hayes (PRA) on MAAP Case: MAAP 02-002-V (CO/CO2 Results from Vogtle RPVRupture Case), November 5, 2002 (copy attached on page 24)6. ASME Steam Tables, Fifth EditionAssumptions:

1. The assumptions and method presented in the WOO documxents (Ref. 1) are valid.2. The containment environment is at 100 % humidity.

3. The temperature and pressure ofconainmenut are within Technical Specification limits when the accident starts.4. The air, steam and hydrogen are released in the same ratio as they exist in containment when venting takes place.The pecet venting is defined as the reduction in the absolute pressure at the time of venting.5. Expected containment failure has been defined as the pressure at which there is a 5% probability of containment

failure, minus 10 psi.6. The SEVERE HYDROGEN CHALLENGE region cannot occur if there is less than 6% hydrogen (wetpercentage),

since a global burn cannot be sustained below this value.Calculation:

To develop CA-3, several of the calculations and the figures for the compuational aid were developed using EXCELspreadsheets.

A. The value of CO and CO2 generaed during 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of corae/oncrete interaction is determined from MAAP A&runs of a severe accident with no containment cooling as recommended in reference

1. This information isshown on page 24.B. The methods used to determine the hydrogen flammability limits in containment are based on the WOOSevere Accident Guidelines (Ref. 1). The equations used are taken directly from these documents and arerepeated below, along with any required design input values.

V12Page 1 of 17Southern Nuclear Operating CornpanySOI AliM Plant: VEGP ITteNt990Re6EACaclios X6CNA1 5I CMPANY Unit: 1&2 Til:NI9-1RvELCluaion SHEET 42 IUNPLANNED increase in Containment Sump, Reactor Coolant Drain Tank(RCDT) or Waste Holdup Tanks (WHT) level of sufficient magnitude toindicate core uncovery.

ANDc. ANY indication from the Containment Challenge Table Cl (above).Containment Operating Deck High Range (RE-O005 or RE-O006):

This radiation monitor reading corresponds to the reflected dose rate from the irradiated fuel inthe RPV with an RPV water level of less than TOAF (18 1'-10" or 63% on Full RangeRVLIS). It is calculated in Attachment E3 of this calculation.

Erratic Source Range Monitor Indication Basis: NEI 99-01 R6, page 74.Explosive mixture inside containment

> 6% by volume hydrogen:

Sheet 23 of VEGP SAMG calculation X6CNA 11 established the 6% by volume hydrogen limit.Pressure

> 14 psig WITH CONTAINMENT CLOSURE established:

NMP-EP-1 10-GL03 (pages 88, 90, 922, & 94) defines CONTAINMENT CLOSURE per"Operating Procedure 14210-1/2, Containment Building Penetrations Verification

-Refueling."

Section 6.0 of 142 10-1/2 lists the acceptance criteria for CONTAINMENT

CLOSURE, amongthem the requirement that >23' of water (EL 21 7"0") is maintained above the RPV flange. Thiscorresponds to the SEP water low level (Design Input #4).During Refueling Operations, fuel is moved between the Containment and the Fuel HandlingBuilding via the Fuel Transfer Tube. The Fuel Transfer Tube will maintain an air-to-air barrierduring refueling operations if either the gate valve is closed or the water level in the refueling cavity is high enough to provide an air-to-air barrier.

Vi12Page 2 of 17Southern Nuclear Operating Company Unit: 1&2 Title: NEI 99-01 Rev 6 EAL Calculations SHEET 43If Containment pressure(PCTMT) exceeds the statichead (AtH) due to thedifference between theTransfer Tube centerline elevation (EL 186"-93/4";

Design Inputs #4 & #5) and P'mthe SEP low operating water level (EL 217'-O";

HDesign Input #4), theTransfer Tube air-to-air

..barrier is not maintained.

',AtH (ft) = 217'-0"-

186'-9.75"

= 30"-2.25"

= -30 ftPctmt (psig) > AtH (ft) x p (Ibm/ft3) x gt (ft/sec2) x I ft2gc (Ibm-ft)/(Ibf-sec

2) 144 in2Pctmt (psig) > 30Oft x 61.551Ibm x 32.2 (ft/sec2) x 1 ft2ft3 32. 2 (Ibm-ft)/(Ibf-sec

2) 144 in2(Design Input #25)> -13 psigPressure

> 52 psig WITH Tech Spec containment integrity intactNMP-EP-1 10-GL03 (page 8) defines CONTAINMENT INTEGRITY as The Primary Containment is OPERABLE per Technical Specification

3. 6.1.1." Tech Spec surveillance requirement 3.6.1.1states "Perform required visual examinations and leakage rate testing except for containment air lock testing, in accordance with the Containment Leakage Rate Testing Program."

TechSpec section 5. 5.17 describes the Containment Leakage Rate Testing Program.

Per Tech SpecBases B3. 6.1, the Containment is designed to contain radioactive material that may be releasedfrom the reactor core following a Design Basis Accident (DBA).Per section 3.1.3.1 of VEGP Design Criteria DC-2101, the mechanical (piping) and electrical penetrations, in conjunction with the carbon steel liner, form a leak-tight barrier.

Thus, thesepenetrations must meet the design accident pressure requirement of section 3. 4.5 of D C-2101,52 psig.The absence of air-to-air containment penetrations during Modes 5 and 6 in VEGP procedure 142 10-1 was confirmed by e-mails from John Stanley (VEGP Operations Outage Manager,"

seeAttachment C5) and Ron Cowen (Westinghouse Site Services Manager; see Attachment C6).

V12Page 3 of 17Piping Penetrations The piping penetrations are listed in drawings 1X4DL 4A0 13, 1X4DL 4A014, 2X4DL 4A0 13, and2X4DL 4A0 14. Cross-sectional views are shown in drawings 1X4DL 4A014 and 2X4DL 4A014.Per section 4.1.2 of specification X4AQIO, these penetrations provide part of the containment boundary.

Section 4.1.3. 3. 2 of this specification directs the user to Attachment 2 for the designtemperature and pressure for these penetrations.

Per Attachment 2 of specification X4AQ1O,the emergency operation design containment pressure is 50 psig.The evaluation in Attachment F of this calc demonstrates that the pipe penetrations should notfail due to a containment pressure of 52 psig.VEGP Condition Report 876376 has been submitted to review and resolve this difference between the Containment and the piping penetration accident design pressure criteria.

Thereare no operability or functionality issues because the peak containment DBA pressure is -37psig (VEGP FSAR Tables 6.2.1-1 & 6.2.1-66).

Electrical Penetrations Per section 3.1.3 of DC-1818, the electrical penetration assemblies shall withstand the pressure, temperature, and environmental conditions resulting from a DBA without exceeding the electrical penetration design leakage rate.Per section E3.6.2 of specification X3AROI-E3, the electrical penetration design leakage rate is0. 01 cc/sec at DBA conditions.

CONTAINMENT CLOSURE no.t established.

Basis: NEI 99-01 Rev 6, page 81.

V12, Page 4 of17Southern Nuclear Design Calculation SPlant: Vogtle Unit: 1&2 ICalculation Number: X6CNA15 ISheet: F-IAttachment F -Evaluation of 52 psig Pressure on Mechanical Penetrations Introduction There is a discrepancy between the DBA Design Pressures for the Containment (52 psig persection 3.4.5 of DC-21 01) and the pipe penetrations (50 psig per Attachment 2 of specification X4AQ1 0).This attachment evaluates the effect of a 52 psig Containment pressure on the pipepenetrations.

Conclusions The compressive and shear loads imposed by a 52 psig Containment pressure on the Unit1&2 pipe penetrations' welds are well below their allowable loads, less than -4% and -30%respectively.

Thus, the pipe penetrations are expected to maintain containment integrity at 52psig.MethodA Type I pipe penetration is shown below:iQlrt & 4N[From 1X4DL4A014

& 2X4DL4A014]

The weakest point of the penetration sleeve is the weld between the penetration sleeve andthe containment liner. If the loads imposed by containment pressure on these welds are lessthan the weld strength, the penetration is expected to maintain containment integrity.

From page 443 of "Strength of Materials":

"The strength of a butt weld is equal to the allowable stress multiplied by the product of the length of the weld times the thickness of the thinnerplate of the joint. The American Welding Society specifies allowable stresses of 20,000 psi intension or compression and 13,600 psi in shear."The specifications for Containment liner welds are likely to be more stringent (i.e., higherallowable stresses) than the values in this textbook.

Using these textbook values isconservative for the purposes of this evaluation:

establishing an allowable limit.

V12, Page 5 of 17Southern Nuclear Design Calculation IPlant: vogtle unit: 1&2 1Calculat°n Nubr: X6CNAI5 sheet: F-2 IAttachment F -Evaluation of 52 psig Pressure on Mechanical Penetrations These allowable stresses are mostlikely specified at standard temperature (68 F or 20 C). The maximum fluidtemperature passing through one ofthese penetrations is 557 F (-290 F).Per VEGP FSAR Table 6.2.1-1, thepeak DBA containment temperature is250 F (-120 C). The yield strength ofsteel decreases with increasing temperature as shown in therepresentative graph to the right.Reducing the above allowable stressesby 15% conservatively addresses theeffect of increased temperature 1,11.00,90,8I-eI-U)0,750 200 400 600Temperature

°CVariation of ultimate strength (Su) and yield strength (Sy)with ratio of operalin temp/Iroom temp (ST/SmT)http:l/www.roymech.co.ukiUsefulTables/Matter/Temperature effects.h~tnl The circumferential weld length (Lw) is calculated as followsLw=ix IDwhereID = Inside diameter of penetration sleeve = OD -2 x tD = OD of penetration sleeve (inches)t = penetration wall thickness (inches)The weld compressive strength (Fc Ibf) is calculates as follows:Fc= [a;cnom X ftemp] X Lw X twhere0c-nom = Nominal allowable compressive stress (20,000 psi)ftemp = Reduction due to increased temperature

= 0.85 = 1 -0.15Lw= Weld length (inches)T = Weld thickness (inches)

= Wall thickness (inches)The weld shear strength (Fs Ibf) is calculates as follows:Fs = [O's-nom X ftemp] X Lw X twhere0s-nora = Nominal allowable shear stress (13,600 psi)fternp = Reduction due to increased temperature

= 0.85 = 1 -0.15Lw= Weld length (inches)T = Weld thickness (inches)

= Wall thickness (inches)

V12, Page 6 of 17Southern Nuclear Design Calculation Plnt Votl Unit: 1& CacltoIume:XCA5sheet:

F-Attachment F -Evaluation of 52 psig Pressure on Mechanical Penetrations The Containment pressure (Pctmt psig) exerts a compressive load (Pc lbf) on the end of thepenetration sleeve. Using the sleeve outside diameter (D in the above figure) maximizes thisload:Pc = Pctmt x H x D2/4The Containment pressure (Pctmt psig) exerts a shear load (Ps Ibf) along the length of thepenetration sleeve. Using the overall sleeve length (L in the above figure) maximizes this load:Ps = Pctmt x H- x D x LEvaluation The effect of a 52 psig Containment pressure on the Unit 1 and Unit 2 pipe penetrations arecalculated in Excel spreadsheets Attachment F1 and Attachment F2.References F1. IX4DL4A0I3, Revision 7, "Containment Building Unit I Containment Wall PipePenetration Design List"F2. IX4DL4A014, Revision 9, "Containment Building Unit I Containment Wall PipePenetration Design List"F3. 2X4DL4A013, Revision 5, "Containment Building Unit 2 Containment Wall PipePenetration Design List"F4. 2X4DL4A014, Revision 4, "Containment Building Unit 2 Containment Wall PipePenetration Design List"F5. Singer, "Strength of Materials,"

second edition, 1962 V12, Page 7 of 17X6CNAI5 ATTACHMENT F SHEET F-4Bornt, ButchFrom: Jani, Yogendra M.Sent: Tuesday, October 14, 2014 4:52 PMTo: Borer, ButchCc: Patel, V. R.; Evans, William P. (SNC Corporate);

Lambert, David Leslie

Subject:

FW: VEGP Pipe Penetration EvalButch,i concur with your methodology used to evaluate 52 psig pressure on MeChanical Penetrations depicted on drawings 1X4DL4AO14

& 2X4DL4A014.

The loads imposed on the weakest point (weld)of penetrations are less than the weld strength.

The penetrations shall exceed the requirements ofASME Section I1I code. So the penetrations are in compliance with specification no. X4AQ10.Therefore, I agree with your conclusion that the pipe penetrations are expected to maintain structural integrity of containment integrity at 52 psig.Thank you,Vogeodra JaniSNCFleet Des -Safety Anl & Mech205.992.5125 office205.410.9806 mobile V12, Page 8 of 17SHEET Fl-iX6CNAI15ATTACHMENT F1Evaluation of 52 psig Pressure on UI Pipe Penetrations L = Overall Length of Penetration Sleeve (inches)D = Penetration Outside Diameter (inches)t = Penetration Sleeve Wall Thickness (inches)ID = Penetration Inside Diameter (inches)ID= D- (2 xt)Lw= Weld Length (inches)Lw= ix IDac = Allowable compressive stress (psi)a3c = a3c-nom X ftemp

= 20,000 psi = Nominal allowable comprssive stressftemp = 0.85 = Reduction due to increased temperature ac = 17,000 psia's =a's =Allowable shear stress (psi)O's-fara X ftemp

= 13,600 psi = Nominal allowable comprssive stressftemp = 0.85 = Reduction due to increased temperature 11,560 psiFc = O'c x Lw x t = Allowable Compressive Load (Ibf)Fs = as x Lw x t = Allowable Shear Load (Ibf)Pctmnt =52 psig = Containment PressureP0 = Compressive Load (lbf)Pc = Pctint x LI x [(D^2)/4]

Ps = Shear Load (Ibf)Ps = Pctmt X (LI X D) X L V12, Page 9 of 17SHEET F1-2X6CNAI5ATTACHMENT F1Evaluation of 52 psig Pressure on U1 Pipe Penetrations Compression LI I= Tp V orVII peerto;dimension B used instead of LType V Penetration:

Per IX4DL4A0 14, there is no dimension L; use B insteadType VII Penetration:

If dimension L not provided on 1X4DL4A013 or IX4DL4A014, use BinsteadPEN # Type L D t ID Lw Fc Pc PclFc1-4 I 48.25 56.000 1.500 53.000 167 4.2E+06 1.3E+05 0.0305 VII 8.250 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.0247-10 I 18.000 18.000 0.500 17.000 53 4.5E+05 1.3E+04 0.02911& 12 III 11.750 12.750 0.375 12.000 38 2.4E+05 6.6E+03 0.02813 III 12.750 12.750 0.375 12.000 38 2.4E+05 6.6E+03 0.02814 VII 15.250 10.750 0.365 10.020 31 2.OE+05 4.7E+03 0.02415 II 15.750 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.02416 -17 V 8.250 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.02418-21 I 37.500 34.000 1.500 31.000 97 2.5E+06 4.7E+04 0.01922 II 15.750 10.750 .0.365 10.020 31 2.0E+05 4.7E+03 0.02423 I 14.930 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.02424 II 15.750 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.02425 V 8.250 18.000 0.500 17.000 53 4.5E+05 1.3E+04 0.02928 &29 II 16.750 18.000 0.500 17.000 53 4.5E+05 1.3E+04 0.02930 & 31 II 18.750 14.000 0.438 13.124 41 3.1E+05 8.0E+03 0.02632 II 15.750 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.02433 II 18.750 14.000 0.438 13.124 41 3.1E+05 8.0E+03 0.02634 & 35 II 19.250 20.000 0.500 19.000 60 5.1E+05 1.6E+04 0.03240 II 16.750 18.000 0.500 17.000 53 4.5E+05 1.3E+04 0.02941 II 15.750 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.02442 II 12.000 10.750 0.365 10.020 31 2.OE+05 4.7E+03 0.02443-46 II 18.750 18.000 0.500 17.000 53 4.5E+05 1.3E+04 0.029V 8.250 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.02448 I 15.750 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.02449 II 15.750 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.02450 I 15.750 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.02451 -55 I 15.750 10.750 0.365 10.020 31 2.OE+05 4.7E+03 0.024 V12, Page 10 of 17SHEET F1-3X6CNA15ATTACHMENT F1Evaluation of 52 psig Pressure on U1 Pipe Penetrations Compression

[111I= TyeV orViipntain dimension B used instead of LType V Penetration:

Per 1X4DL4AO014, there is no dimension L; use B insteadType VII Penetration:

if dimension L not provided on 1X4DL4A013 or 1X4DL4AO14, use BinsteadPEN # Type JLJD~ t IDJ__ Fc_ Pc JPcIFc56 I__44.750134.000I1.500 31.0001 97 2.5E+06 4.7E+04 0.01957 & 58 I 32.000 24.000 1.000 22.000 69 1 .2E+06 2.4E+04 0.02059 & 60 I 28.000 26.000 1.000 24.000 75 1 .3E+06 2.8E+I04 0.02261 V 8.250 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.02462 &63 II 15.750 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.02464 VI 13.250 10.750 0.365 10.020 31 2.0E+05 4.7E+i03 0.02466 V j8.250 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.02467 III 12.750 12.750 0.375 12.000 38 2.4E+05 6.6E+03 0.02868 II 16.000 18.000 0.500 17.000 53 4.5E+05 1.3E+04 0.02969 -73 III 12.750 12.750 0.375 12.000 38 2.4E+05 6.6E+03 0.02875 V 8.250 10.750 0.365 10.020 131 2.0E+'05 4.7E+03 0.02476 V 8.250 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.02477 & 78 II 15.750 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.02479 II 12.000 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.02480 II 16.000 18.000 0.500 17.000 53 4.5E+05 1.3E+'04 0.02981 II 12.000 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.02482 V 8.250 18.000 0.500 17.000 53 4.5E+05 1.3E+04 0.02983 & 84 10.000 24.000 0.500 23.000 72 6.1E+05 2.4E+04 0.03885 III 12.750 12.750 0.375 12.000 38 2.4E+05 6.6E+03 0.02886 III 23.670 12.750 0.375 12.000 38 2.4E+05 6.6E+03 0.02887 II 16.000 18.000 0.500 17.000 53 4.5E+05 1.3E+04 0.02988 VII 15.250 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.02490 VII 8.250 10.750 0.365 10.020 31 2.0E+05 4.7E+'03 0.02491 -98 II 19.750 18.000 0.750 16.500 52 6.6E+05 1.3E+04 0.020100 9.250 4.500 0.237 4.026 13 5.1E+04 8.3E+02 0.016101 -104 I 20.550 18.000 0.750 16.500 52 6.6E+05 1.3E+04 0.020Maximum PclFc = 0.038 V12, Page 11 of 17SHEET FI-4X6CNAI15ATTACHMENT F1Evaluation of 52 psig Pressure on UI Pipe Penetrations Shear[111= Type V or VII penetration; dimension B used instead of LType V Penetration:

Per IX4DL4A0 14, there is no dimension L; use B insteadType VII Penetration:

If dimension L not provided on 1X4DL4A013 or 1X4DL4A014, use BinsteadPEN # IType L 0 t ID Lw Fs Ps JPs/Fs1 -4 I 48.250 56.000 1.5001 53.0001167 2.9E+06 4.4E+05 0.1535 VII 8.250 10.750 0.365 10.020 31.5 1.3E+05 1.4E+04 0.1097- 10 I 18.000 18.000 0.500 17.000 53.4 3.1E+05 5.3E+04 0.17111& 12 III 11.750 12.750 0.375 12.000 37.7 1.6E+05 2.4E+04 0.15013 III 12.750 12.750 0.375 12.000 37.7 1.6E+05 2.7E+04 0.16314 VII 15.250 10.750 0.365 10.020 31.5 1.3E+05 2.7E+04 0.20215 II 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.20816- 17 V 8.250 10.750 0.365 10.020 31.5 1.3E+05 1.4E+04 0.10918-21 I 37.500 34.000 1.500 31.000 97.4 1.7E+06 2.1E+05 0.12322 II 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.20823 I 14.930 10.750 0.365 10.020 31.5 1.3E+05 2.6E+04 0.19724 II 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.20825 V 8.250 18.000 0.500 17.000 53.4 3.1E+05 2.4E+04 0.07928 & 29 II 16.750 18.000 0.500 17.000 53.4 3.1E+05 4.9E+04 0.16030 & 31 II 18.750 14.000 0.438 13.124 41.2 2.1E+05 4.3E+04 0.20532 II 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.20833 II 18.750 14.000 0.438 13.124 41.2 2.1E+05 4.3E+04 0.20534 & 35 II 19.250 20.000 0.500 19.000 59.7 3.5E+05 6.3E+04 0.18240 II 16.750 18.000 0.500 17.000 53.4 3.1E+05 4.9E+04 0.16041 II 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.20842 II 12.000 10.750 0.365 10.020 31.5 1.3E+05 2.1E+04 0.15943-46 II 18.750 18.000 0.500 17.000 53.4 3.1E+05 5.5E+04 0.17947 V 8.250 10.750 0.365 10.020 31.5 1.3E+05 1.4E+04 0.10948 I 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.20849 II 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.20850 I 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.20851 -55 I 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.208 V12, Page 12 of 17SHEET Fl-5X6CNA1 5ATTACHMENT FlEvaluation of 52 psig Pressure on UI Pipe Penetrations ShearI ... I=Type V or VII penetration; dimension B used instead of LType V Penetration:

Per IX4DL4A0 14, there is no dimension L; use B insteadType VII Penetration:

If dimension L not provided on IX4DL4A013 or 1X4DL4A014, use BinsteadPEN # Type L D t ID Lw Fs Ps P5/F556 I 44.750 34.000 1.500 31.000 97.4 1.7E+06 2.5E+05 0.14757 & 58 I 32.000 24.000 1.000 22.000 69.1 8.0E+05 1.3E+05 0.15759 & 60 I 28.000 26.000 1.000 24.000 75.4 8.7E+05 1.2E+05 0.13661 V o8.25'0 10.750 0.365 10.020 31.5 1.3E+05 1.4E+04 0.10962 &63 II 15.750 10.750 0.365 10.020 31.5 i.3E+05 2.8E+04 0.20864 VI 13.250 10.750 0.365 10.020 31.5 1.3E+05 2.35+04 0.175,,66 ...IV 8.250; 10.750 0.365 10.020 31.5 1.3E+05 1.4E+04 0.10967 III 12.750 12.750 0.375 12.000 37.7 1.6E+05 2.7E+04 0.16368 II 16.000 18.000 0.500 17.000 53.4 3.1E+05 4.7E+04 0.15269 -73 III 12.750 12.750 0.375 12.000 37.7 1.6E+05 2.7E+04 0.163:75: V °8,250 10.750 0.365 10.020 31.5 1.3E+05 1.4E+04 0.109:V 8.250 10.750 0.365 10.020 31.5 1.3E+05 1.4E+04 0.10977 & 78 II 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.20879 II 12.000 10.750 0.365 10.020 31.5 1.3E+05 2.IE+04 0.15980 II 16.000 18.000 0.500 17.000 53.4 3.IE+05 4.7E+04 0.15281 II 12.000 10.750 0.365 10.020 31.5 1.3E+05 2.1E+04 0.159182 ....8.250 18.000 0.500 17.000 53.4 3.1E+05 2.4E+04 0.07983 & 84 10.000 24.000 0.500 23.000 72.3 4.2E+05 3.9E+04 0.09485 III 12.750 12.750 0.375 12.000 37.7 1.6E+05 2.7E+04 0.16386 III 23.670 12.750 0.375 12.000 37.7 1.6E+05 4.9E+-04 0.30287 II 16.000 18.000 0.500 17.000 53.4 3.IE+05 4.7E+04 0.15288: VII i,.250' 10.750 0.365 10.020 31.5 1.3E+05 2.7E+04 0.202'90 " VII 8.250 10.750 0.365 10.020 31.5 1.3E+05 1.4E+04 0.10991 -98 II 19.750 18.000 0.750 16.500 51.8 4.5E+05 5.8E+04 0.129100 9.250 4.500 0.237 4.026 12.6 3.5E+04 6.8E+03 0.196101,-104 I 20.550 18.000 0.750 16.500 51.8 4.5E+05 6.0E+04 0.134Maximum PslFs = 0.302 V12, Page 13 of 17X6CNA15 ATTACHMENT F2 SHEET F2-1Evaluation of 52 psig Pressure on U2 Pipe Penetrations L = Overall Length of Penetration Sleeve (inches)D = Penetration Outside Diameter (inches)t = Penetration Sleeve Wall Thickness (inches)ID = Penetration Inside Diameter (inches)I D= D-(2xt)Lw= Weld Length (inches)Lw= fixiD-c= Allowable compressive stress (psi)c'c = O'c-nom X ftempac-norn = 20,000 psi = Nominal allowable comprssive stressftermp = 0.85 = Reduction due to increased temPerature O-c = 17,000 psi = Allowable compressive stressas = Allowable shear stress (psi)a's = a's-nom X ftempa'c-nomn

= ,I13,600 psi = Nominal allowable comprssive stressftemp = 0.85 = RedUction due to increased temperature O's = 11,560 psi = Allowable shear stressFc = ac x Lw x t = Allowable Compressive Load (Ibf)Fs = as x Lw x t = Allowable Shear Load (Ibf)Pctmnt = 52 psig = Containment PressurePc = Compressive Load (lbf)Pc = Pctmt x H x [(D^2)/4]

Ps = Shear Load (Ibf)Ps = Pctmt X (H- X D) X L V12, Page 14 of 17SHEET F2-2X6CNA15SATTACHMENT F2Evaluation of 52 psig Pressure on U2 Pipe Penetrations Compression I I= Type V or VII penetration; dimension B used instead of LiType V Penetration:

Per 2X4DL4A0 14, there is no dimension L; use B insteadType VII Penetration:

If dimension L not provided on 2X4DL4A013 or 2X4DL4A014, use BinsteadPEN # Type L D t ID Lw FcPc PclFc1 -4 I 48.25 56.000 1.500 53.000 167 4,2E+06 1.3E+05 0.0305 VII 8.250 10.750 0.365 10.020 31 2,0E+05 4,7E+03 0.0247- 10 I 18.000 18.000 0.500 17.000 53 4.5E+05 1,3E+04 0.02911& 12 III 11.750 12.750 0.375 12.000 38 2,4E+05 6,6E+03 0.02813 III 12.750 12.750 0.375 12.000 38 2.4E+05 6.6E+03 0.02814 VII 15.250 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.02415 II 15.750 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.02416 -17 V 8.250 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.02418-21 I 37.500 34.000 1.500 31.000 97 2.5E+06 4,7E+04 0.01922 II 15.750 10.750 0.365 10.020 31 2,0E+05 4,7E+03 0.02423 I 14.930 10.750 0.365 10.020 31 2.0E+05 4,7E+03 0.02424 II 15.750 10.750 0.365 10.020 31 2,0E+05 4,7E+03 0.02425 ,V 8.250 18.000 0.500 17.000 53 4.5E+05 1,3E+04 0.02928 & 29 II 16.750 18.000 0.500 17.000 53 4.5E+05 1,3E+04 0.02930 & 31 II 18.750 14.000 0.438 13.124 41 3.1E+05 8,0E+03 0.02632 II 15.750 10.750 0.365 10.020 31 2,0E+05 4,7E+03 0.02433 II 18.750 14.000 0.438 13.124 41 3,1E+05 8,0E+03 0.02634 & 35 II 19.250 20.000 0.500 19.000 60 5.IE+05 1,6E+04 0.03240 II 16.750 18.000 0.500 17.000 53 4,5E+05 1,3E+04 0.02941 II 15.750 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.02442 II 12.000 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.02443 -46 II 18.750 18.000 0.500 17.000 53 4.5E+05 1.3E+04 0.029"47 V 8.250 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.02448 I 15.750 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.02449 II 15.750 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.02450 I 15.750 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.02451 -55 I 15.750 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.024 V12, Page 15 of 17SHEET F2-3X6CNA15SATTACHMENT F2Evaluation of 52 psig Pressure on U2 Pipe Penetrations Compression

~=TpeorI=

pentraIon dimension Bused instead of LType V Penetration:

Per 2X4DL4A0 14, there is no dimension L; use B insteadType VII Penetration:

If dimension L not provided on 2X4DL4A013 or 2X4DL4A014, use BinsteadPEN # Type L D t ID LFcPc Pc/Fc56 I 44.750 34.000 1.500 31.000 97 2.5E+06 4.7E+04 0.01957 & 58 I 32.000 24.000 1.000 22.000 69 1.2E+06 2.4E+04 0.02059 & 60 1 28.000 26.000 1.000 24.000 75 1 .3E+06 2.8E+04 0.02261 V 8.250 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.02462 &63 II 15.750 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.02464 VI 13.250 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.02466 V 8.250 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0,02467 III 12,750 12,750 0.375 12,000 38 2,4E+05 6.6E+03 0,02868 II 16.000 18.000 0.500 17.000 53 4.5E+05 1.3E+04 0.02969-73 III 12.750 12.750 0.375 12.000 38 2.4E+05 6.6E+03 0.02875 V 8.250 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.02476 V 8.250 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.02477 &78 II 15.750 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.02479 II 12,000 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.02480 II 16.000 18.000 0.500 17.000 53 4.5E+05 1.3E+04 0.02981 II 12.000 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.02482 V 8.250 18.000 0.500 17.000 53 4.5E+05 1.3E+04 0.02983 &84 10.000 24.000 0.500 23.000 72 6.IE+05 2.4E+04 0.03885 III 12.750 12.750 0.375 12.000 38 2.4E+05 6.6E+03 0.02886 III 23.670 12.750 0.375 12.000 38 2.4E+05 6.6E+03 0.02887 II 16.000 18.000 0.500 17.000 53 4.5E+05 1.3E+04 0.02988 VII 15.250 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.02490 VII 8.250 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.02491 -98 II 19.750 18.000 0.750 16.500 52 6.6E+05 1.3E+04 0.020100 9.250 4.500 0.237 4.026 13 5.1E+04 8.3E+02 0.016101 -104 I 20.550 18.000 0.750 16.500 52 6.6E+05 1.3E+04 0.020Maximum PdIFc = 0.038 V12, Page 16 of 17SHEET F2-4X6CNA15SATTACHMENT F2Evaluation of 52 psig Pressure on U2 Pipe Penetrations ShearType V orVillpntain dimension B used instead of LType V Penetration:

Per 2X4DL4A014, there is no dimension L; use B insteadType VII Penetration:

If dimension L not provided on 2X4DL4A013 or 2X4DL4A014, use BinsteadPEN # Type L D t ID Lw Fs Ps PslFs1 -4 I 48.250 56.000 1.500 53.000 167 2.9E+06 4.4E+05 0.1535 VII 8.250 10.750 0.365 10.020 31.5 1,3E+05 1.4E+04 0.1097-10 I 18.000 18.000 0.500 17.000 53.4 3.1E+05 5.3E+04 0.17111& 12 III 11.750 12.750 0.375 12.000 37.7 1.6E+05 2.4E+04 0.15013 III 12.750 12.750 0.375 12.000 37.7 1.6E+05 2.7E+04 0.16314 VII 15.250 10.750 0.365 10.020 31.5 1.3E+05 2.7E+04 0.20215 II 15.750 10.750 !0.365 10.020 31.5 1.3E+05 2.8E+04 0.20816-17 V 8.250 10.750 0.365 10.020 31.5 1.3E+05 1.4E+04 0.10918-21 1 37.500 34.000 1.500 31.000 97.4 1.7E+06 2.IE+05 0.12322 II 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.20823 I 14.930 10.750 0.365 10.020 31.5 1.3E+05 2.6E+04 0.19724 II 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.20825 V 8.250 18.000 0.500 17.000 53.4 3.1E+05 2.4E+04 0.07928 &29 II 16.750 18.000 0.500 17.000 53.4 3.1E+05 4.9E+04 0.16030 & 31 II 18.750 14.000 0.438 13.124 41.2 2.1E+05 4.3E+04 0.20532 II 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.20833 II 18.750 14.000 0.438 13.124 41.2 2.IE+05 4.3E+04 0.20534 & 35 II 19.250 20.000 0.500 19.000 59.7 3.5E+05 6.3E+04 0.18240 II 16.750 18.000 0.500 17.000 53.4 3.IE+05 4.9E+04 0.16041 II. 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.20842 II 12.000 10.750 0.365 10.020 31.5 1.3E+05 2.1E+04 0.15943-46 II 18.750 18.000 0.500 17.000 53.4 3.1E+05 5.5E+04 0.17947 V 8.250 10.750 0.365 10.020 31.5 1.3E+05 1.4E+04 0.10948 I 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.20849 II 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.20850 I 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.20851-55 I 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.208 V12, Page 17 of 17SHEET F2-5X6CNA1 5ATTACHMENT F2Evaluation of 52 psi9 Pressure on U2 Pipe Penetrations Shear[111= Type V or VII penetration; dimension B used instead of LType V Penetration:

Per 2X4DL4A014, there is no dimension L; use B insteadType VII Penetration:

If dimension L not provided on 2X4DL4A013 or 2X4DL4A014, use BinsteadPEN # Type L D t ID L FsPs PslFs56 I 44.750 34.000 1.500 31.000 97.4 1.7E+06 2.5E+05 0.14757 & 58 I 32.000 24.000 1.000 22.000 69.1 8.0E+05 1.3E+05 0,15759 & 60 I 28.000 26.000 1.000 24.000 75.4 8.7E+05 1.2E+05 0.13661 V 8.250 10.750 0.365 10.020 31.5 1.3E+05 1.4E+04 0.10962 &63 II 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.20864 VI 13.250 10.750 0.365 10.020 31.5 1.3E+05 2.3E+04 0.17566 V 8.250 10.750 0.365 10.020 31.5 1.3E+05 1.4E+04 0.10967 III 12.750 12.750 0.375 12.000 37.7 1.6E+05 2.7E+04 0.16368 II 16.000 18.000 0.500 17.000 53.4 3.1E+05 4.7E+04 0.15269-73 III 12.750 12.750 0,375 12.000 37.7 1.6E+05 2.7E+04 0.16375 V 8.250 10.750 0.365 10.020 31.5 1.3E+05 1.4E+04 0.10976 V 8.250, 10.750 0.365 10.020 31.5 1.3E+05 1.4E+04 0.10977 &78 II 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.20879 II 12.000 10.750 0.365 10.020 31.5 1.3E+05 2.1E+04 0.15980 II 16.000 18.000 0.500 17.000 53.4 3.1E+05 4.7E+04 0.15281 II 12.000 10.750 0.365 10.020 31.5 1.3E+05 2.1E+04 0.15982 V 8.250 18.000 0.500 17.000 53.4 3.1E+05 2.4E+04 0.07983 & 84 10.000 24.000 0.500 23.000 72.3 4.2E+05 3.9E+04 0.09485 III 12.750 12.750 0.375 12.000 37.7 1.6E+05 2.7E+04 0,16386 III 23.670 12.750 0.375 12.000 37.7 1.6E=+05 4.9E=+04 0.30287 II 16.000 18.000 0.500 17.000 53.4 3.1E+05 4.7E+04 0.15288 VII 15.250 10.750 0.365 10.020 31.5 1.3E=+05 2.7E+04 0.202,90, VII 8.250 10.750 0.365 10.020 31.5 1.3E+05 1.4E+04 0.10991 -98 II 19.750 18.000 0.750 16.500 51.8 4.5E+05 5.8E+04 0.129100 9.250 4.500 0.237 4.026 12.6 3.5E+04 6.8E+03 0.196101 -104 I 20.550 18.000 0.750 16.500 51.8 4.5E+05 6.0E+-04 0.134Maximum PslFs 0.302 I-+/- 4.4.4 .J~XH~_ ~ Th.7r~.jVi3Page 1 of 1T<-H........

.....--I ALVI W. VOCTHE P1AJhC.;a i i a i V1 4Page 1 of 2provided to prevent explosive concentrations of hydrogen during battery charging.

The ventilation system shall be adequate to maintain the hydrogen concentration below 2 % in accordance with IEEE 484.*Each room shall be provided with a shower and an eye-wash basin in case personnel come in contact with battery acid.*To prevent possible hydrogen explosion, switches and receptacles shall not be locatedinside battery rooms.3.2 POWER GENERATION DESIGN BASESNone3.3 MAJOR COMPONENT DESIGN BASES3.3.1 The Class IE dc 125-V dc system and all equipment are classified as Seismic Category 1and Safety Class 1E. The system equipment shall be capable of continuous operation between 100 and 140 V dc, except vital ac bus inverters, the reactor trip switchgear

control, residual heat removal (RHR) isolation valve inverters, and the turbine-driven auxiliary feedwater pump control may be allowed to operate over a dc input voltagerange of 105 to 14OV dc.The 25 kVA inverters provide power to RIHR isolation valves. The breakers for thesevalves are located in "Trip" position during normal plant operation; they do not operateduring the duty cycle time of the battery, or for any LOCA/LOOP or SBO copingscenario.

3.3.2 Reqiuirements for Batteries 3.3.2.1 Battery size shall be determined in accordance with the method indicated in IEEE 485.3.3.2.2 Each Class 1E battery shall have sufficient capacity to independently supply the requiredloads for a loss-of-coolant

accident, loss of offsite power, or main steam line break for aduration of 2.75 hr. Each Class 1E battery shall have sufficient capacity to independently supply the required loads for a station blackout for a duration of 4 hr.3.3.2.3 Initial battery capacity shall be 25 % greater than required according to the calculation method indicated in IEEE 485 to allow for aging and extend the time interval for batteryreplacement as required by the battery replacement criteria of IEEE 450.3.3.2.4 Batteries shall be sized to provide their required output at 70°F.3.3.2.5 A margin of 10 % load growth shall be initially included in the sizing of each battery.DC-i1806 6VR16VER 13 V14Page 2 of 23.3.3.7 A dc ammeter, dc voltmeter, dc overvoltage relay, ac power "on" light, and acundervoltage relay shall be provided on each charger.

The relays shall alarm in thecontrol room.3.3.3.8 The chargers shall be suitable for parallel operation so that each subsystem's redundant charger can be manually put into operation in conjunction with the normal charger torecharge a discharged battery in a shorter amount of time (if allowed by themanufacturer).

3.3.3.9 The battery charger shall prevent the charger from becoming a load on the battery due toa power feedback during loss of ac power to the chargers.

3.3.3.10 ac breakers shall be provided to protect the charger from internal faults and to isolate thecharger from the ac source.3.3.3.11 dc breakers shall be provided to protect the battery and charger from internal chargerfaults and to isolate the charger from the dc system.3.3.4 Requirements for 125-V dc Metal-Enclosed Switchgear 3.3.4.1 There is one dc switchgear lineup for each subsystem.

It shall be connected to the battery,the normal battery charger, and the redundant battery charger associated with that train.The switchgear shall feed MCCs, dc distribution panels, and the inverter for the vitalinstrumentation system (DC- 1807).3.3.4.2 The switchgear air circuit breakers shall be equipped with direct-acting, dual-magnetic, overcurrent tripping devices providing adjustable overcurrent and short-circuit protection.

3.3.4.3 The 125-V dc switchgear breakers shall serve as a means for energizing and deenergizing power sources and loads connected to the 125-V dc switchgear bus. The switchgear shallalso provide suitable protection for the loads during overload and short-circuit conditions.

3.3.4.4 Trains C and D shall each provide 125-V dc power to an associated 480-V, 3-phaseinverter for RHR isolation valves.3.3.4.5 The vital ac bus inverters may be allowed to operate over a dc input voltage range of 105to 140 V dc. The dc feeder cables shall be designed to maintain a minimum of 105 V dcduring the entire battery load profile.3.3.4.6 The RHR isolation valve inverters may be allowed to operate over a range of 10._5 to140 V dc. The dc feeder cables to the RHR isolation valve inverters shall be designed tomaintain a minimum of 105 V dc at the RH-R isolation valve inverters.

3.3.5 125-V dc MCCs3.3.5.1 One MCC shall be provided for each train A, B, and C subsystem.

It shall be connected to the switchgear associated with that subsystem.

The MCC shall feed motor-operated DC-1806 8VR18VER 13 X6CNA1 5Attachment L

C2I.l~at~inn fnr I:W.I I1SHEET L-6 v15Page 1 of 3CALC NO. SNC024-CALC-004 VEGP DETERMINATION OFEMERGENCY ACTION LEVEL REV. 0I0 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 6.2-3] are provided below in Table 5-1. Thesesource values are scaled to develop the emergency action levels for initiating condition E-HU1.Table 5-1 Technical Specification (Neutron

+Gamma) Dose Rate Limits for HI-STORM100 and HI-TRAC 125Number of Technical Specification LoatonjMeasurements j Limit (mrem/hr)

HI-TRAC 1.25 __________

Side -Mid -height 4 472.7Top 1 4 j102.4HI-STORM 100 ____ _____Side -60 inches below mid-height 4 87Side -Mid -height 4 88.9Side -60 inches above mid-height 4 54.8Top -Center of lid 1 24.5Top -Radially centered 4 29.2Inlet duct 4 178.8Outlet duct 4 64.5 X6CNA1 5Attachment LENERCON Calculation for E-HU1SHEET L-7 v15Page 2 of 3CALC NO. SNC024-CALC-004 VEGP DETERMINATION OF ____________

EMERGENCY ACTION LEVEL REV. 00 E N E R C 0 N FOR INITIATING CONDITION E- -__________

PAGE NO. Page 7of 86.0 Methodology The "on-contact" dose rates from the technical specification for the HI STORM-i100 casksystem are scaled by a factor of 2, as specified in NEI 99-01 Rev. 6 [1], for use in initiating condition E-HUI.

X6CNA1 5Attachment LFNFRf.ON

('nlrmiitinn fnr F-HI- 11SHEET L-8V1Page 3 of 3CALC NO. SNC024-CALC-004 VEGP DETERMINATION OF -____________

EMERGENCY ACTION LEVEL REV. 00 E N E R C 0 N FOR INITIATING CONDITION E-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 LimitsTechnical LoainSpecification Scaling Calculated Value EALLoainLimit Factor (mrem/hr)

(mrem/hr)

_________________________

(mrem/hr) j____ _______________

HI-TRAC 125___ _____Side -Mid -height [ 472.7 2 1 945.4 [ 950Top 102.4 2 j 204.8 [ 200HI-STORM 100Side -60 inches below mid-height 87 2 174 170Side -Mid -height 88.9 2 177.8 180Side -60 inches above mid-height 54.8 2 109.6 110Top -Center of lid 24.5 2 49 50Top -Radially centered 29.2 2 58.4 60Inlet duct 178.8 2 357.6 360Outlet duct 64.5 2 129 1308.0 Computer SoftwareMicrosoft WORD 2013 is used in this calculation for basic multiplication.

V1 6Page 1 of 4Approved ByJ. B. StanleyEffective Date7/25/12~Voqgtle Electric Generating F-0 CRITICAL SAFETY FUNCTION SiF- 0.2CORE COOLINGPatProcedure versionPat19200-C 24.2Page Number'ATUS TREES5olSheet 1 of 1-4 GO TO11221-V~GO TO19221-CVLI5 FULL NOGE GREATERtHAN 41% ,E8S* .GO TO19222-C11.FI GO TO19222Z-CtVUI FULL NiGE GREATERTHAN 41% YE.GO TO* 19223-V~GO TOrYNAMIC Ha 9,Z-INGE Nho -4 RCPh-3 RCPh-2RCP E-1IRCF* 19223-CGI CSA TvPrinted February 2, 2016 at 16:09 Vi16Page 2 of 4Approved By Procedure VesoJ. B. Stanley Vogtle Electric Generating P..lantng 19200-C Veso24.2Effective Date Page Number7/25/12 F-0 CRITICAL SAFETY FUNCTION STATUS TREES6of1 Sheet 1 of 1F- 0.3HEAT SINKIKTOTAL AVAILABLE FEEDWATER FLOWTO SGs GREATERiTHAN 570 GPMNOYESNARROW RANGELEVEL IN AT LEASTrONE SG GREATERTHAN 10% (32%)GO TO1 9231 -CGO TO19232-CGO TO1 9233-C NOPRESSURE IN ALL SGsLESS THAN 1240 PSIG YELESS THAN 82% YESNOPRESSURE IN ALL SGsLESS THAN 1180 PSIGYESGO TO1 9234-CGO TO19235-CNARROW RANGE NOLEVEL IN ALL SGsGREATER THAN10% (32%) YESLSATPrinted February 2, 2016 at 16:09 V1 6Page 3 of 4Approved By Procedure VersionJ. B. Stanley Vogtle Electric Generating Plant 19200-C 24.2Effective Date Page Number7/25/12 F-0 CRITICAL SAFETY FUNCTION STATUS TREES7of1 Sheet 1 of 1F- 0.4INTEGRITY GO TO19241-C---- .) GO TOI 19241-CALL RCS WR COLD LEGTEMPERATURES GREATER THAN 2950F~QGO TO19242-C~*CSF SATTEMPERATURE DECREASE IN ALL NORCS COLD LEGS LESS1000F IN THEILAST iYES!60 MINUTESJGOTOI,11 19241-CALL RCS WR COLD LEG NOTEMPERATURES GREATERTHAN 265°F YESRCS PRESSURE LESS NOTHAN COLDOVERPRESSURE LIMIT 465 PSIG YES,IRCS WR COLD LEG NOTEMPERATURE GREATER THAN 2200F YESLL* (I'\ GO TO" "" 19242-CO CSF SAT-*CSF SATPrinted February 2, 2016 at 16:09 Vi16Page 4 of 4Approved By .....Procedure VersionJ. B. StanI~yVogtle Electric Generating Plant '19200-C 2.Effe tiv Datan ey P... ..age... :.. .... .... ... ...... NumbeEfeciv DteF-0 CRITICAL SAFETY FUNCTION STATUS TREES 9ag ofmb1r7/25/129of1 Sheet 1 of 1F- 0.5CONTAINMENT GO TO19251-C* )GO TOi 19251-CN.E N oPGO TOp 19252-CICONTAINMENT BUMP LEVEL LESSTHAN 196 iNCHESGO TO.......19253.CIbCSF SATPrinted February 2, 2016 at 16:09 V1 7Page 1 of 3Approved By p, 4Pocdr VersionW. L. Burmeister

... Vogtle Electric Generating Plant IS55039-c~rcdr 3.2Effective Date P~age Number0510712013 ISEISMIC MONITORING INSTRUMENTATION SYSTEM I 6 of 94.2 NORMAL OPERATION NONE4.3 NON-PERIODIC OPERATIONS NOTEThis subsection shall be initiated by 50022-C "Seismic Event Plan" orSupervisor's direction.

14.3.1 Retrieving Seismic DataNOTESKey Number 1-OP3-10 is in C&T. LIa. Verify the Event alarm on the Condor Control Unit screen isRED. Elb. Obtain the event charts and graphs from each recorder.

El4.3.2 Retrieving Seismic Data At ETNA (River Intake)a. Hook up a laptop to the uplink cable. Elb. Power on computer, THEN click on the ALTUS Quick Talkicon. Elc. In the ALTUS Status window, verify the alarm has beentriggered.

Eld. Save the file to a floppy disk. El(1) Click on the EVTI folder in the ALTUS directory window El(2) Highlight the EVTI file associated with the recordedevent El(3) Click on the Retrieve File button El(4) Download the file to a floppy disk ElPrinted February 15, 2016 at 16:59 V17Page 2 of 3Seismic Event Plan 50022-CVOGTLE Version 14.0Unit C Page 7of 244.0 INSTRUCTION 4.1 IDENTIFICATION OF SEISMIC EVENT1. The following indicators are available for determining whether or not seismicevent has occurred:

a. The Event Alarm on the Condor Control Unit screen is RED.b. The National Earthquake Information Center, located in Denver,Colorado, Telephone (303) 273-8500, confirms that an earthquake hasoccurred; (Must be called when any of the above indications of anearthquake is received) and to initiate this procedure if earthquake isconfirmed.

2. IF an earthquake is sensibly detected by control room personnel, initiate thisprocedure.

3. IF AOP 18036-C, Seismic Event, is in effect, then initiate this procedure.

Printed February 15, 2016 at 17:02 V1 7Page 3 of 3Seismic Event 18036-CVOGTLE Version 11Unit C Page 4of 9PURPOSEThe purpose of this procedure is to provide operator response following a seismic event and toinitiate an engineering analysis to determine the severity of the event.SYMPTOMS* Actuation of seismic monitor alarm.* Actuation of seismic instrumentation.

• Effects of earthquake heard or felt.MAJOR ACTIONS* Evaluate effects of seismic event.* Determine if shutdown of the units is required.

Printed February 15, 2016 at 17:05 VI18Page 1 of 5Southern Nuclear Design Calculation IPlant: Vogte Unit: 1&2 Icalculation Number: X6CNAI4 Isheet:;46 Miscellaneous Design Inputs21. Iodine boiling point = 184 C = -363 F

Reference:

Page B-I, "C3RC Handbook of Chemistry

& Physics"22. Density of Refueling Cavity and Spent Fuel Pool Water @ 130 F = 61.55 Ibm/cu ft

Reference:

See Attachment C32.23. Density of C3VCS letdown flow = 0.99 g/cc (Attachment C2)

Reference:

The density is used to convert the letdown activity from p.(Ci/g to ltCi/cc, whichare the units used by the C3VCS letdown rad monitor RE-48000 (Design Input #1 &Attachment CS5). Based on at-power CVCS letdown parameters from the Unit 1 and 2 IPCs(Attachment C35), the average temperature and pressure at the radiation measurement location are 98.5 F and 385 psig.24. Average Decay Gamma Energies for RE-48000 principle isotopes (Attachment C38)I rIsotopeAverageGammaEnergy(MeV)

Reference:

Brookhaven National Laboratory NationalNuclear Data Center decay data(http://www.orau

.or qlptp/PTP%20Libraryllibrary/DOE/bnl/nu clidedata/table.htm)

Copies of web pages in Attachment C81-131 0.3821-132 2.201-133 0.607I-134 2.501-135 1.55Co-580.975 Co-60 2.51Cs-134 1.55Cs-136 2.12Cs-i137 0.565Cs-1382.31Cs-138 2.31 Vi18Page 2 of 5Southern Nuclear Design Calculation SPlant: Vogtle Unit: 1&2 ICalculation Number: X6CNA14 ISheet: 61Recognition Category S: System Malfunctions Notice of Unusual EventSU4: Fuel Clad Degradation.

Operating Mode Applicability:

Power Operation (Mode 1)Startup (Mode 2)Hot Standby (Mode 3)Hot Shutdown (Mode 4)1 OR2Emergency Action Levels:SU4 EALI: CVCS Letdown radiation monitor RE-48000 reading greater than 5 pCi/ccindicating fuel clad degradation greater than Technical specification allowable limits.There are two Technical Specification limits on RCS coolant activity:

• SR 3.4.16.1:

Gross specific activity

< pCi/gm* SR 3.4.16.2:

Dose Equivalent 1-131 (DE 1-131) < 1.0 !iCi/gPer section B.3.4.16, page B3.4.16-2 of VEGP Tech Spec Bases, noble gasactivity in the reactor coolant assumes 1% failed fuel, which closely equalsthe LCO limit of 1 00/1s pCi/gm for gross specific activity.

The EAL threshold will be calculated for each Tech Spec limit condition.

Per pages 12 and 13of X6AZ01 A, theprinciple isotopesdetected by RE-48000 are 1-131, 1-133, Co-58, Co-60,Cs-134, and Cs-137.However, per SectionB-12-3-2 and FigureB-12-2 of 1X6AZ01-10004 & 2X6AZ01-10004, RE-48000 willdetect gammas ofenergies down to-0.1 MeV.St1 __ _ "__I-, -.. I _ _ _ _i c -i ..L =t 4 ..II; P.mIENKR4Y It ,VgL= t.VIFigure B-12-2Thus the other I, Co, and Cs isotopes listed in FSAR Table 11.1-2 should beincluded if their average decay gamma energies exceed 0.1 MeV.

V1 8Page 3 of 5Southern Nuclear Design Calculation iPlant: Vogtle U nit: 1&2 ICalculation Number: X6CNAI4 Sheet: 62Per LTR-CRA-06-179 attached to WEC-SNC letter GP-18006, the pre-MURPU coolant activities may be adjusted upward 2% to account for theincrease in core thermal power from 3565 MWt to 3636 MWt. Thus, the Coand Cs MURPU 1% defect activity are equal to their pre-MURPU 1% Defectactivities multiplied by 1.02.The Co and Cs activities corresponding to the 1.0 DE 1-131 TechSpec limit are the products of their MURPU 1% defect activities and theratio of the 1-131 DE 1-131 concentration to its equilibrium concentration (0.74/2.91).

The activities, expressed in j!iCi/g are summed and then multiplied by theCVCS letdown flow density (0.99 g/cc) to convert them to The EAL threshold is the minimum of the 1% Defect and the 1 .0 DE I-131 activities.

1.0 MURPU Pre-MURPU DE I-131 1% Defect 1% DefectIsotope Coolant Coolant CoolantActivity Activity ActivityI-131 0.74 2.91 ______I-132 0.75 2.96 ______I-133 1.41 5.561-134 0.18 0.69 ______I-135 0.69 2.72 ______Co-58 3.89E-03 1 .53E-02 1 .50E-02Co-60 4.93E-04 1 .94E-03 1 .90E-03Cs-134 5.97E-01 2.35 2.3Cs-I136 7.52E-01 2.96 2.9Cs-137 3.89E-01 1.53 1.5Total = 5.5 21.7 ptCi/gTotal = 5.5 21.5 i.LCi/ccCVCS Letdown Density =0.99g/ccSGiven the RG 1.97 R2 required system accuracy (Acceptance Criterion 3),the threshold is rounded down from 5.5 to 5 jltCi/cc.

NOTE: SU4 EAL2 not determined in this calculation.

V1 8Southern Nuclear Design Calculation Page 4 of 5SPlant: Vogtle Unit: 1&2 Calculation Number: X6CNA14 Sheet: C5-1Attachment C5 -VEGP 1&2 CVCS Letdown Radiation Monitor (RE-48000)

ReadingsU 1.... ... .aIIII-~' I~~-:-~a..

V1 8Southern Nuclear Design Calculation Page 5 of 5SPlant: Vogtle Unit: 1&2 Calculation Number: X6CNA14 Sheet: C5-2Attachment C5 -VEGP 1&2 CVCS Letdown Radiation Monitor (RE-48000)

Readings*]W11o=I~IMY Iin'~vu l~u~r ~.~tImP~.ii I~' ~'~IWE ~'~~jL 11U WOWW4m~ ~

V19Page 1 of 3RCS Specific Activity3.4.163.4 REACTOR COOLANT SYSTEM (RCS)3.4.16 RCS Specific ActivityLCO 3.4.16APPLICABILITY:

The specific activity of the reactor coolant shall be within limits.MODES 1 and 2,MODE 3 with RCS average temperature (Tavg) > 500°F.ACTIONS--------------------------

INlJLCO 3.0.4c is applicable.

I------------------

---CONDITION REQUIRED ACTION COMPLETION TIMEA. DOSE EQUIVALENT A.1 Verify DOSE Once per4 hoursI-131 > 1.0 p.Ci/gm.

EQUIVALENT I-131within the acceptable region of Figure 3.4.16-1.

ANDA.2 Restore DOSE 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />sEQUIVALENT I-131 towithin limit.B. Gross specific activity of B.1 Perform SR 3.4.16.2.

4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />sthe reactor coolant notwithin limit. AND8.2 Be in MODE 3 with 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />sTavg < 500°F.(continued)

Vogtle Units 1 and 23.4.16-1Amendment No. 137 (Unit 1)Amendment No. 116 (Unit 2)

V1 9Page 2 of 3RCS Specific Activity3.4.16ACTIONS (continued)

________________

__________

CONDITION REQUIRED ACTION COMPLETION TIMEC. Required Action and C.1 Be in MODE 3 with 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />sassociated Completion Tavg < 500°F.Time of Condition A notmet.O_.RDOSE EQUIVALENT 1-131 in theunacceptable region ofFigure 3.4.16-1.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.16.1 Verify reactor coolant gross specific In accordance withactivity_

100/I. !iCi/gm.

the Surveillance Frequency ControlProgramSR 3.4.16.2

---

--NOTE- --- --Only required to be performed in MODE 1.Verify reactor coolant DOSE EQUIVALENT I-131 In accordance withspecific activity

< 1.0 ,.tCi/gm, the Surveillance Frequency ControlProgramANDBetween 2 and6 hours after aTHERMAL POWERchange of _> 15% RTPwithin a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period(continued)

Vogtle Units 1 and 23.4.16-2Amendment No. 158 (Unit 1)Amendment No. 140 (Unit 2)

V1 9Page 3 of 3RCS Specific Activity3.4.16250IU-I.20015010050PERCENT OF RATED THERMAL POWERFIGURE 3.4.16-1REACTOR COOLANT DOSE EQUIVALENT I-131 REACTOR COOLANT SPECIFIC ACTIVITYLIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANTSPECIFIC ACITVITY

>1 mCi/gram DOSE EQUIVALENT 1-131Vogtle Units 1 and 23.4.16-4Amendment No. 96 (Unit 1)Amendment No. 74 (Unit 2)

V20Page 1 of 1RCS Operational LEAKAGE3.4.133.4 REACTOR COOLANT SYSTEM (RCS)3.4.13 RCS Operational LEAKAGELCO 3.4.13RCS operational LEAKAGE shall be limited to:a. No pressure boundary LEAKAGE;b. I1 gpm unidentified LEAKAGE;c I1 p dniidLAAE nd. 150 galosper idaytprimdLAryGE toscndar EKG hog none steam generator (SG).APPLICABILITY:

MODES 1, 2, 3, and 4.ACTIONS__________________

___CONDITION REQUIRED ACTION COMPLETION TIMEA. RCS operational A.1I Reduce LEAKAGE to 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />sLEAKAGE not within within limits.limits for reasons otherthan pressure boundaryLEAKAGE or primary tosecondary LEAKAGE.B. Required Action and B.1 Be in MODE 3. 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />sassociated Completion Time of Condition A not ANDmet.B.2 Be in MODE 5. 36 hour4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />sO__RPressure boundaryLEAKAGE exists.ORPrimary to secondary LEAKAGE not withinlimit.Vogtle Units 1 and 23.4.13-1Amendment No. 144 (Unit 1)Amendment No. 124 (Unit 2)

V21IPage 1 of 2Approved ByPoede VrinJ. B. Stanley Vogtle Electric Generating Plant 19200cdr 24.2i~Effective Date -0CTIASAEYFNTOSAUSRES Page Number7/25/12 F 0 C I CA SA E Y F N T O ST T S R ES9 of 11Sheet 1 of 1F- 0.5CONTAINMENT GOaTO19251-C.-'- PRESURELS u ==s4j TANji sl jI* a O T19251-(;IJ i .AT LEAST ONESCONTAINMENT SSPRAY PUMPSRUNNINGNOYES*egoGO TO* il ) 19261-CIGO TO19252-Cb/T> GO TO: ......1 9 2 6 3 -CCSP SAT,-r'nneu rebruary iZUll at 14:zz V21Page 2 of 2S0uthern Nuclear Operating CompanyA~rlN Plant: VEGP ! "X6CNA1 5ISUHda Unit: 1&2 Title: NEI 99-01 Rev 6 EAL Calculations SHEET 43If Containment pressure(PCTMT) exceeds the statichead (AH) dlue to thedifference between theTransfer Tube centerline elevation (EL 186"-93/4";

Design In puts #4 & #5) and PT~the SFP low operating water level (EL 21 Design Input #4), theTransfer Tube air-to-air barrier is not maintained."

AIH (ft) = 217'-0"-

186'-9.75"

= 30"-2.25"

= -30 ft(psig) > AH (ft) x p (Ibn/ft3) x g1 (ft/sec2) x 1 ft2go (Ibm -ft)/(lIbf-sec

2) 144 in2Pctmt (psig) > 30 ft x 61.55 Ibm x 32.2 (ft/sec2) x 1 ft2t332.2 (Ibm-ft)/(Ibf-sec

2) 144 in2(Design Input #25)Petmi > '43 psigPressure

> 52 psig WITH Tech Spec containment integrity intactNMP-EP-110-GL03 (page 8) defines CONTAINMENT INTEGRITY as The Primary Containment is OPERABLE per Technical Specification

3. 6.1.1." Tech Spec surveillance requirement

3. 6.1.1states "Perform required visual examinations and leakage rate testing except for containment air lock testing, in accordance with the Containment Leakage Rate Testing Program."

TechSpec section 5. 5.17 describes the Containment Leakage Rate Testing Program.

Per Tech SpecBases B3. 6.1, the Containment is designed to contain radioactive material that may be releasedfrom the reactor core following a Design Basis Accident (DBA).Per section 3.1.3.1 of VEGP Design Criteria DC-2 101, the mechanical (piping) and electrical penetrations, in conjunction with the carbon steel liner, form a leak-tight barrier.

Thus, thesepenetrations must meet the design accident pressure requirement of section 3. 4.5 of DC-2 101,52 psig.The absence of air-to-air containment penetrations during Modes 5 and 6 in VEGP procedure 142 10-1 was confirmed by e-mails from John Stanley (VEGP Operations Outage Manager; seeAttachment CS) and Ron Cowen (Westinghouse Site Services Manager; see Attachment C6).

V9Page 1 of 5Southern Nuclear Operating CompanyI rlll Plant: VEGP TteNE990Re6EACacliosI X6CNA1 5Unit: 1&2 TteNE9-0Re6ELCacliosSHEET 10Volume fraction above operating deck = 0.771

Reference:

Table 6.5.2-1, VEGP FSAR Revision 19 (February 2014)8. Containment liner: 1/4"4 carbon steel

Reference:

VEGP FSAR sections 1.2.5, 6.2.7.2,

& 6.5.3.1 and drawings 1X2D01A001

&2X2D01A001 Reactor Coolant System Parameters

9. Reactor Pressure Vessel & RCS Piping Dimensions Parameter Value Reference RPV Inside Diameter 173" VEGP FSAR Table 5.3.3-1Hot Leg centerline elevation 187'-0" AX4DR023, 1X4DL4A017-1,

&(76% RVLIS) 2X4DL4A01 7-1Cold Leg centerline elevation Hot Leg Nozzle Bottom 185'-91/2' AX4DR023Top of Active Fuel 181'-10" AX4DR023(63% RVLIS)Cold Leg Pipe ID 27%" 1X4DL4A017-1

& 2X4DL4A017-1 Hot Leg Pipe ID 29" 1X4DL4A017-1

& 2X4DL4A017-1 RCS Coolant Parameters Parameter Value Reference Full Power Tavg 588.4 °F Table 2-1, page 2-3, WCAP-16736-P VEGP FSAR Table 15.0.3-3RCS operating pressure 2250 psiaFull power coolant mass 2.53E+08 g Page 3 of LTR-CRA-06-1 79 attachedto WEC-SNC letter GP-1 8006 andTable 7.8-3 of WCAP-16736-P 10.11. Fuel Assembly outside dimensions

= 8.424" x 8.424"

Reference:

1 X6AN09-1 0000-2 & 2X6AN09-1 0000-012. Core effective diameter

= 132.7 inches x 1 foot/12 inches = 11.06 ft

Reference:

Table 5-1, page 5-4, 1/2X6AA10-00095 Source Terms V9Page 2 of 5Southern Nuclear Operating CompanySOI AmiE Plant: VEGP ITitle: NEI 99-01 Rev 6 EAL Calculations X6CNA15~I Unit: 1&2 SHEET 37CA1: Loss of RPV inventory.

Operating Mode Applicability:

Emergency Action Levels:Cold Shutdown, Refueling 1 OR21. Loss of RPV inventory as indicated be level less than elevation 185'-10" (73% on FullRange RVLIS).The RPV water level elevations corresponding to the RCS loop piping bottom IDsare found as follows:Dimension IElevation Loop Centerline Elevation 1 87'-00"Cold LegInside Diameter 27.5"1/2AxID 13.75"Bottom ID = Centerline

-(1/2Ax ID) 185'-1 0.25"Hot LegInside Diameter 29.0"1/2Ax ID 14.5"Bottom ID = Centerline

-(A x ID) 185'-9.5" The dimensions and elevations are taken from Design Input #9. The RPV waterlevel elevation corresponding to the Bottom ID of the RCS piping is ~185'1O".

Because the core barrel is a right circular

cylinder, the RVLIS indication corresponding to the above RPV water level can be determined by linearlyinterpolating between the TOAF (EL 181'-10" or 63% RVLIS) and the CL and HLcenterline elevation (EL 187"0O" or 76% RVLIS):

V9Page 3 of 5Southern Nuclear Operating CornpanyouIu A i= Plant: VEGP ITteNE990Re6EACacliosX6CNA1 5 II ULOMPA Unit: 1&2 Til:NI9-Rv6ELCluaion SHEET 38VEGP RVLIS Indication vs. RPV Water Level Elevation

} .i181 182 F 18 8 8 8 88RPV Water Level Elevation (feet)The RPV water level elevation corresponding to the Bottom ID is 185'-10" or~73% on Full Range RVLIS.2. a. RPV level cannot be monitored for 15 minutes or longerANDb. UNPLANNED increase in Containment sump, Reactor Coolant Drain Tank (RCDT)or Waste Holdup Tank (WHT) levels due to a loss of RPV inventory.

V9Page 4 of 5Southern Nuclear Operating Company4xnlmM. Plant: VEGP ITitle: NEI 99-01 Rev 6 EAL Calculations X 6CNA15 ISO mUTH t Unit: 1&2 S HEET 39CSI: Loss of RPV inventory affecting core decay heat removal capability.

Operating Mode Applicability:

Emergency Action Levels:Cold Shutdown, Refueling 1 OR20OR31. a. CONTAINMENT CLOSURE not established ANDb. RPV water level less than 1 85'-4" [6" below Bottom ID of loop] (72% on Full RangeRVLIS).The RPV water level elevations corresponding to 6" below the cold leg (CL) andhot leg (HL) bottom IDs are found as follows:Dimension Elevation Loop Centerline Elevation 1 87'-00"Cold LegInside Diameter 27.5"% x ID 13.75"Bottom ID = Centerline

-(1/2 x ID) 185'-10.25" 6" Below CL Bottom ID 1 85'-4.25" Hot LegInside Diameter 29.0"% xID 14.5"Bottom ID = Centerline

-(1/2 x ID) 185'-9.5" 6" Below HL Bottom ID 1 85'-3.5"The dimensions and elevations are taken from Design Input #9. The elevation corresponding to 6" below the Bottom ID of the RCS piping is ~185'4".Because the core barrel is a right circular

cylinder, the RVLIS indication corresponding to the above RPV water level can be determined by linearlyinterpolating between the TOAF (EL 181 '-10" or 63% RVLIS) and the CL and HLcenterline elevation (EL 187"-0" or 76% RVLIS):

V9Page 5 of 5Southern Nuclear Operating CompanyPlnt: VEGP Title: NEI 99-01 Rev 6 EAL CalculationsX6N1 mpw Unit: 1&2 I SHEET 40VEGP RVLIS Indication vs. RPV Water Level Elevation 621 "B~owRCS-

__ ___.... .... ..... i .jPiping BottomU)181 182 183 184 185 188 187 188RPV Water Level Elevation (feet)The RPV water level elevation corresponding to 6" below the Bottom ID is185'-4" or -72% on Full Range RVLIS.2. a. CONTAINMENT CLOSURE established ANDb. RPV level less than 181'-1 0" ITOAF] (63% on Full Range RVLIS).3. a. RPV level cannot be monitored for 30 minutes or longerANDb. Core uncovery is indicated by ANY of the following:

RE-005 O..R 006 > 40 REM/hrErratic Source Range monitor indication UNPLANNED increase in Containment Sump, Reactor CoolantDrain Tank (RCDT) or Waste Holdup Tanks (WHT) levels ofsufficient magnitude to indicate core uncovery Vl 0VEGP-FSAR-1 1Pae1o511.2.1.3 Eqluipment DesignThe LWPS equipment design parameters are provided in table 11.2.1-2.

The seismic design classification and safety classification for the LWPS components andstructures are listed in table 3.2.2-1.

Safety class designations are also indicated on the LWPSpiping and instrumentation

diagram, drawings 1X4DB 124, 1X4DB 125, 1X4DB 126, 1X4D B127,AX4DB1 24-2, AX4DB 124-3, AX4DB1 24-4, and AX4DB 124-5.11.2.1.4 Reference

1. U.S. Nuclear Regulatory Commission, "Calculation of Releases from Pressurized WaterReactors,"

NUREG-0017, April 1976.11.2.2 SYSTEM DESCRIPTIONS The liquid waste processing system (LWPS) collects and processes potentially radioactive wastes for recycling or release to the environment.

Provisions are made to sample and analyzefluids before discharge.

Based on the laboratory

analysis, these wastes are either retained forfurther processing or released under controlled conditions through the cooling water system,which dilutes the discharge flow. A permanent record of liquid releases is provided by analysesof known volumes of effluent.

The radioactive liquid discharged from the reactor coolant system (RCS) is processed by theradwaste processing facility systems and may be discharged or recycled.

The LWPS is arranged to recycle reactor grade water if desired.

This is implemented by thesegqregqation of equipment drains and waste streams to prevent intermixingq of liquid wastes.The LWPS can be divided into the following subsystems:

A. Reactor Coolant Drain Tank (RCDT) Subsystem This portion of the LWPS collects nonaerated, reactor grade effluent fromsources inside the containment.

B. Drain Channel AThis portion of the LWPS collects

aerated, reactor grade effluent that can berecycled.

C. Drain Channel BThis portion of the LWPS processes all effluent that is not suitable for recycling.

D. Radwaste Processing Facility Demineralizers The radwaste processing facility demineralizer systems consist of portabledemineralizers installed in subterranean enclosures inside the radwasteprocessing facility.

The radwaste processing facility is described in paragraph 11.4.2.4.

The radwaste processing facility demineralizers can be aligned toprocess any of the three waste drain streams.E. The radwaste processing facility filtration system consists of a portable, vendorsupplied system located within a shielded area inside the radwaste processing facility.

The filtration system associated tanks, pumps, accumulator, piping,valves, and controls located within a shielded area inside the radwaste11.2-411.2-4REV 13 4106 Vl0aPage 2 of 5VEGP-FSAR-1 1processing facility.

The peripheral equipment is located adjacent to the filterassembly.

The filter system can be aligned to process any of the three wastedrain streams.

Details of this equipment are shown on drawing AX4DB1 24-1.In addition, the LWPS provides capability for handling and storage of spent ion exchangeresins.The LWPS does not include provisions for processing secondary system wastes. Secondary system effluent is handled by the steam generator blowdown processing system (SGBPS),

asdescribed in subsection 10.4.8, and by the turbine building drain system. Estimated releasesfrom these systems are discussed in subsection 11.2.3. The LWPS design, which segregates primary and secondary wastes, minimizes the amount of water that must be processed bydischarging low activity wastes directly, where permissible, with no treatment.

Instrumentation and controls necessary for the operation of the LWPS are located on a controlboard in the auxiliary building.

Any alarm on this control board (except for the waste processing holdup control panel) is relayed to the main control board in the control room.The LWPS piping and instrumentation diagrams are shown in drawings

1X4DB124, 1X4DB125,

1X4DB126, 1X4DB127, AX4DB124-1, AX4DB124-2, AX4DB124-3, AX4DB124-4, andAX4DB1 24-5 and process flow diagram for the LWPS is shown on figure 11.2.2-1.

Table11.2.1-1 lists the assumptions regarding flows and activity levels that were used in preparation of table 11.2.1-3, which gives nuclide concentrations for key locations within the LWPS asshown on figure 11.2.2-1.

The process flow data is calculated using the data in table 11.2.1-1, the flow paths indicated on figure 11.2.2-1, realistic primary coolant activity levels from section11.1, and decontamination factors as given in reference 1 of subsection 11.2.1.11.2.2.1 Reactor Coolant Drain Tank Subsystem IRecyclable reactor grade effluents enter this subsystem from valve leakoffs, reactor coolantIpump No. 2 seal leakoffs, reactor vessel flange leakoff, and other deaerated, tritiated waterIsources inside the containment.

Connections are provided for draining the RCS loops and thesafety injection system (SIS) accumulators and for cooling the pressurizer relief tank. Inaddition, refueling canal drains can be routed to the refueling water storage tank using theRCDT pumps.The RCDT contents are continuously recirculated through the RCDT heat exchanger tomaintain the desired temperature.

Level is prevented from varying significantly by a controlvalve which automatically opens a path from the recirculation line to the BRS when normal tanklevel is exceeded.

The RCDT is also connected to the gaseous waste processing system(GWPS) vent header. Hydrogen gas bottles connected to the RCDT ensure a hydrogenblanket.

Maintaining a constant level minimizes the amount of gas sent to the GWPS andminimizes the amount of hydrogen used. Provisions for sampling the gas are provided.

Details of the RCDT subsystem are shown on drawing 1X4DB127.

A separate RCDTsubsystem is provided for each of the two units.11.2.2.2 Drain Channel A Subsystem

Aereated, tritiated liquid enters drain channel A through lines connected to the waste holduptank. Sources of this aerated liquid are as follows:A. Accumulator drainage (via RCDT pump suction).

11.2-511.2-5REV 13 4/06 V10oPage 3 of 5VEGP-FSAR-11I B. Sample room sink drains (excess primary sample volume only).C. Ion exchanger, filter, pump, and other equipment drains.The containment sump or auxiliary building sump may be directed to the waste holdup tank orthe floor drain tank for processing as necessary.

The collected aerated drainage is pumped or flows to the waste holdup tank prior to processing through the radwaste processing facility filtration system and/or the radwaste processing facilitydemineralizers before reuse or discharge.

Details of this equipment are shown on drawingsAX4DB1 24-2, AX4DB1 24-3, AX4DB1 24-4, and AX4DB1 24-5.The basic composition of the liquid collected in the waste holdup tank is boric acid and waterwith some radioactivity.

A separate drain channel A subsystem is provided for each of the two units. Details are shownon drawings 1X4DB124 and 1X4DB127.

Table 11.2.1-1 lists the estimated flows entering thewaste holdup tank.11.2.2.3 Drain Channel B Subsystem Drain channel B is provided to collect and process nonreactor grade liquid wastes. Theseinclude:* Wastes from floor drains.* Equipment drains containing nonreactor grade water.* Laundry and hot shower drains.* Other nonreactor grade sources.Drain channel B is comprised of three drain subchannels, each associated with one of thefollowing tanks.A. Laundry and Hot Shower TankThe laundry and hot shower tank is provided to collect and process wasteeffluents from the plant laundry and personnel decontamination showers andhand sinks.Laundry and hot shower drains normally need no treatment for removal ofradioactivity.

This water is transferred to a waste monitor tank through thelaundry and hot shower tank filter for eventual discharge.

If sample analysisindicates that decontamination is necessary, the water can be directed throughthe Unit 1 or Unit 2 waste monitor tank demineralizer or the radwaste processing facility for cleanup.The laundry and hot shower tank and filter are shared by the two units. Detailsof this portion of the LWPS are shown on drawing 1X4DB126.

Table 11.2.1-1lists estimated flows entering the laundry and hot shower tank.B. Floor Drain TankWater may enter the floor drain tank from system leaks inside the containment through the containment sump, from system leaks in the auxiliary buildingthrough auxiliary building sumps and the floor drains, and floor drains in the11.2-611.2-6REV 13 4/06 v10oPage 4 of 5VEGP-FSAR-1 1radwaste facilities.

Sources of water to the containment sump and auxiliary building sumps and floor drains are the following:

1. Fan cooler leaks.2. Secondary side steam and feedwater leaks.3. Primary side process leaks.4. Decontamination water.The containment sump or auxiliary building sumps may be directed to the wasteholdup tank.Another source of water to the floor drain tank is the chemical laboratory drains.Excess nonreactor grade samples that are not chemically contaminated andlaboratory equipment rinse water are drained to the floor drain tank.The contents of the floor drain tank are processed through the radwasteprocessing facility demineralizers and/or the radwaste processing facility filtration system and then pumped to a waste monitor tank for ultimate discharge.

If the activity in the floor drain tank liquid is such that the discharge limits cannotbe met without cleanup, the liquid can be processed by the waste monitor tankdemineralizer, the radwaste processing facility demineralizers, or the radwasteprocessing facility filtration system.A separate floor drain tank and associated equipment are provided for each ofthe two units. Details of this portion of the LWPS are shown on drawing1X4DB126.

Table 11.2.1-1 lists the estimated flows entering the floor drain tank.C. Chemical Drain TankLaboratory samples which contain reagent chemicals (and possibly tritiated liquid) are discarded through a sample room sink which drains to the chemicaldrain tank. Chemical drains requiring radwaste processing are sent to the solidwaste management system or may be processed through the radwasteprocessing facility demineralizers and/or the radwaste processing facility filtration system.The chemical drain tank and associated equipment are shared by Units 1 and 2.Details of this portion of the LWPS are shown on drawing 1X4DB125.

Table11.2.1-1 lists the estimated flow directed to the chemical drain tank.Any liquids released to the environment by the LWPS are first directed to a waste monitor tank.Before releasing the contents of a waste monitor tank, a sample is taken for analysis.

Thefindings are logged, and, if the activity level is within acceptable limits, the tank contents arereleased to the discharge canal. The discharge valve is interlocked with a process radiation monitor and closes automatically when the radioactivity concentration in the liquid discharge exceeds a preset limit. The radiation element is located upstream of the discharge valve at adistance sufficient to close the valve before passing the fluid that activated the detector tripsignal. The isolation valve also blocks flow if sufficient dilution water is not available.

Theradiation monitor is described in section 11.5. A permanent record of the radioactive releasesis provided by a sample analysis of the known volumes of waste effluent released.

Liquidwaste discharge flow and volume are also recorded.

If the monitor tank contents are not acceptable for discharge, the fluid can be held for a time toallow activity to decay to acceptable levels, or it can be further processed by the waste monitor11.2-711.2-7REV 13 4/06 V10Page 5 of 5VEGP-FSAR-11I H. Waste Monitor Tank PumpsTwo pumps are provided for each unit. One pump is used for each monitor tankto discharge water from the LWPS or for recycling if further processing isrequired.

The pump may also be used for circulating the water in the waste monitor tank toobtain uniform tank contents, and therefore a representative sample, beforedischarge.

These pumps can be throttled to achieve the desired discharge rate.I. Auxiliary Waste Monitor Tank PumpsTwo pumps are provided.

They are installed in Unit 2 but serve both units. Onepump is used for each auxiliary waste monitor tank to discharge water fromLWPS or for recycling if further processing is required.

A mixer may be used forcirculating the water in the auxiliary waste monitor tank to obtain uniform tankcontents, thereby assuring a representative sample is acquired prior to discharge of the tank contents.

The pumps can be throttled to achieve the desireddischarge rate.11.2.2.6.2 TanksA. Reactor Coolant Drain TankOne tank is provided for each unit. The purpose of the RCDT is to collectleakoff-type drains inside the containment at a central collection point for furtherdisposition through a single penetration via the RCDT pumps. The tank providessurge volume and net positive suction head (NPSH) to the pumps.Only water which can be directed to the boron recycle holdup tanks enters theRCDT. The water is compatible with reactor coolant and does not containdissolved air during normal plant operation, by engineering design.A constant level is maintained in the tank to minimize the amount of gas sent tothe GWPS and also to minimize the amount of hydrogen cover gas required.

The level is maintained by one continuously running pump and by a control valvein the discharge line. This valve operates on a signal from a level controller tolimit the flow out of the system. The remainder of the flow is recirculated to thetank.Continuous flow is maintained through the heat exchanger in order to preventloss of pump NPSH resulting from a sudden inflow of hot liquid into the RCDT.B. Waste Holdup TankOne atmospheric pressure tank is provided for each unit to collect:1. Equipment drains.2. Valve and pump seal leakoffs (outside the containment).

3. Boron recycle holdup tank overflows.

4. Other water from tritiated, aerated sources.The tank size is adequate to accommodate 11 days of expected influent duringnormal operation.

C. Waste Evaporator Condensate Tank11.2-1111.2-11REV 13 4/06 ViiPage 1 of 3Southern Nuclear Operating CompanysmrllllM~LPlant:

VEGP Title: NEI 99-01 Rev 6 EAL Calculations I 6CNA15¢m Unit: 1&2 SHEET 42UNPLANNED increase in Containment Sump, Reactor Coolant Drain Tank(RCDT) or Waste Holdup Tanks (WHT) level of sufficient magnitude toindicate core uncovery.

ANDc. ANY indication from the Containment Challenge Table Cl (above).Containment Operating Deck High Range (RE-O005 or RE-O006):

This radiation monitor reading corresponds to the reflected dose rate from the irradiated fuel inthe RPV with an RPV water level of less than TOAF (18 1'-10" or 63% on Full RangeRVLIS). It is calculated in Attachment E3 of this calculation.

Erratic Source Range Monitor Indication Basis: NEI 99-01 R6, page 74.Explosive mixture inside containment

> 6% by volume hydrogen:

Sheet 23 of VEGP SAMG calculation X6CNA1 1 established the 6% by volume hydrogen limit.Pressure

> 14 psig WITH CONTAINMENT CLOSURE established:

NMP-EP-1 10-GL03 (pages 88, 90, 922, & 94) defines CONTAINMENT CLOSURE per"Operating Procedure 142 10-1/2, Containment Building Penetrations Verification

-Refueling."

Section 6.0 of 142 10-1/2 lists the acceptance criteria for CONTAINMENT

CLOSURE, amongthem the requirement that >23' of water (EL 21 7'-0") is maintained above the RPV flange. Thiscorresponds to the SEP water low level (Design Input #4).During Refueling Operations, fuel is moved between the Containment and the Fuel HandlingBuilding via the Fuel Transfer Tube. The Fuel Transfer Tube will maintain an air-to-air barrierduring refueling operations if either the gate valve is closed or the water level in the refueling cavity is high enough to provide an air-to-air barrier.

VllPage 2 of 3Southern Nuclear Operating CornpanyAOm~I 4 Plant: VEGP Title: NEI 99-01 Rev 6 EAL Calculations I X6CNA1 5I MV Unit: 1&2 I SHEET 53The results of the Loss of Clad FP Barrier setpoint calculations in Attachments H-3 and 13 aresummarized below. Given the system accuracy

-a factor of two over the operating range -thethreshold is rounded off to two significant figures.Unit Calculated Threshold Rounded-Off Threshold (R EM/h r) (m RE M/h r)VEGP 1 1.31E+04 1.3E+07VEGP 2 1.49E+04 1.5E+07Containment Barrier Potential Loss Threshold 4.BContainment Hydrogen concentration greater than 6%.The existence of an explosive mixture means, at a minimum, that the containment atmospheric hydrogen concentration is sufficient to support a hydrogen burn (i.e., at the lower deflagration limit). A hydrogen burn will raise containment pressure and could result in collateral equipment damage leading to a loss of containment integrity.

It therefore represents a potential loss of thecontainment barrier.Sheet 23 of VEGP SAMG calculation X6CNA II established the 6% by volume hydrogen limit.

VllPage 3 of 3Desis,,C.lulation

-Nuclear Southern Conmpay Services Aim~Prjc:Vogtl lei Gener~atig Plat Ci.No. X6CNAn 1 5SSubJectflitle:

Severe Accident Management Guideline (SAMG) Calculations Sheet 23 of 167CA-3 HYDROGEN FLAMMABILITY IN CONTAINMENT Determined Values: See attached graphs.Guidelines:

SAG-2, 3, 7, SCG-3

References:

1. MUHP-23 10, WOG Severe Accident Management Guidance (Background Document) and MUHP-23 15 I //sWOO Severe Accident Management Guidance.,

Rev. 12. EpRI TR-101 869, Severe Acodient Managemnent Guidance Technical Basis Report, Volume 2: The Physicsof Accident Progression

3. FSAR:a. Secion 6.2.1.5.2

c. Figure 6.2.1-1 INb. Table 6.2.5-6 d. Figur 6.2.1-44. Technical Specifications:

a. Section 3.6.1.4b. Section 3.6.1.5 1/5. Memo from Roger Hayes (PRA) on MAAP Case: MAAP 02-002-V (CO/CO2 Results from Vogtle RPVRupture Case), November 5, 2002 (copy attached on page 24)6. ASME Steam Tables, Fifth EditionAssumptions:

1. The assumptions and method presented in the WOO documxents (Ref. 1) are valid.2. The containment environment is at 100 % humidity.

3. The temperature and pressure ofconainmenut are within Technical Specification limits when the accident starts.4. The air, steam and hydrogen are released in the same ratio as they exist in containment when venting takes place.The pecet venting is defined as the reduction in the absolute pressure at the time of venting.5. Expected containment failure has been defined as the pressure at which there is a 5% probability of containment

failure, minus 10 psi.6. The SEVERE HYDROGEN CHALLENGE region cannot occur if there is less than 6% hydrogen (wetpercentage),

since a global burn cannot be sustained below this value.Calculation:

To develop CA-3, several of the calculations and the figures for the compuational aid were developed using EXCELspreadsheets.

A. The value of CO and CO2 generaed during 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of corae/oncrete interaction is determined from MAAP A&runs of a severe accident with no containment cooling as recommended in reference

1. This information isshown on page 24.B. The methods used to determine the hydrogen flammability limits in containment are based on the WOOSevere Accident Guidelines (Ref. 1). The equations used are taken directly from these documents and arerepeated below, along with any required design input values.

V12Page 1 of 17Southern Nuclear Operating CornpanySOI AliM Plant: VEGP ITteNt990Re6EACaclios X6CNA1 5I CMPANY Unit: 1&2 Til:NI9-1RvELCluaion SHEET 42 IUNPLANNED increase in Containment Sump, Reactor Coolant Drain Tank(RCDT) or Waste Holdup Tanks (WHT) level of sufficient magnitude toindicate core uncovery.

ANDc. ANY indication from the Containment Challenge Table Cl (above).Containment Operating Deck High Range (RE-O005 or RE-O006):

This radiation monitor reading corresponds to the reflected dose rate from the irradiated fuel inthe RPV with an RPV water level of less than TOAF (18 1'-10" or 63% on Full RangeRVLIS). It is calculated in Attachment E3 of this calculation.

Erratic Source Range Monitor Indication Basis: NEI 99-01 R6, page 74.Explosive mixture inside containment

> 6% by volume hydrogen:

Sheet 23 of VEGP SAMG calculation X6CNA 11 established the 6% by volume hydrogen limit.Pressure

> 14 psig WITH CONTAINMENT CLOSURE established:

NMP-EP-1 10-GL03 (pages 88, 90, 922, & 94) defines CONTAINMENT CLOSURE per"Operating Procedure 14210-1/2, Containment Building Penetrations Verification

-Refueling."

Section 6.0 of 142 10-1/2 lists the acceptance criteria for CONTAINMENT

CLOSURE, amongthem the requirement that >23' of water (EL 21 7"0") is maintained above the RPV flange. Thiscorresponds to the SEP water low level (Design Input #4).During Refueling Operations, fuel is moved between the Containment and the Fuel HandlingBuilding via the Fuel Transfer Tube. The Fuel Transfer Tube will maintain an air-to-air barrierduring refueling operations if either the gate valve is closed or the water level in the refueling cavity is high enough to provide an air-to-air barrier.

Vi12Page 2 of 17Southern Nuclear Operating Company Unit: 1&2 Title: NEI 99-01 Rev 6 EAL Calculations SHEET 43If Containment pressure(PCTMT) exceeds the statichead (AtH) due to thedifference between theTransfer Tube centerline elevation (EL 186"-93/4";

Design Inputs #4 & #5) and P'mthe SEP low operating water level (EL 217'-O";

HDesign Input #4), theTransfer Tube air-to-air

..barrier is not maintained.

',AtH (ft) = 217'-0"-

186'-9.75"

= 30"-2.25"

= -30 ftPctmt (psig) > AtH (ft) x p (Ibm/ft3) x gt (ft/sec2) x I ft2gc (Ibm-ft)/(Ibf-sec

2) 144 in2Pctmt (psig) > 30Oft x 61.551Ibm x 32.2 (ft/sec2) x 1 ft2ft3 32. 2 (Ibm-ft)/(Ibf-sec

2) 144 in2(Design Input #25)> -13 psigPressure

> 52 psig WITH Tech Spec containment integrity intactNMP-EP-1 10-GL03 (page 8) defines CONTAINMENT INTEGRITY as The Primary Containment is OPERABLE per Technical Specification

3. 6.1.1." Tech Spec surveillance requirement 3.6.1.1states "Perform required visual examinations and leakage rate testing except for containment air lock testing, in accordance with the Containment Leakage Rate Testing Program."

TechSpec section 5. 5.17 describes the Containment Leakage Rate Testing Program.

Per Tech SpecBases B3. 6.1, the Containment is designed to contain radioactive material that may be releasedfrom the reactor core following a Design Basis Accident (DBA).Per section 3.1.3.1 of VEGP Design Criteria DC-2101, the mechanical (piping) and electrical penetrations, in conjunction with the carbon steel liner, form a leak-tight barrier.

Thus, thesepenetrations must meet the design accident pressure requirement of section 3. 4.5 of D C-2101,52 psig.The absence of air-to-air containment penetrations during Modes 5 and 6 in VEGP procedure 142 10-1 was confirmed by e-mails from John Stanley (VEGP Operations Outage Manager,"

seeAttachment C5) and Ron Cowen (Westinghouse Site Services Manager; see Attachment C6).

V12Page 3 of 17Piping Penetrations The piping penetrations are listed in drawings 1X4DL 4A0 13, 1X4DL 4A014, 2X4DL 4A0 13, and2X4DL 4A0 14. Cross-sectional views are shown in drawings 1X4DL 4A014 and 2X4DL 4A014.Per section 4.1.2 of specification X4AQIO, these penetrations provide part of the containment boundary.

Section 4.1.3. 3. 2 of this specification directs the user to Attachment 2 for the designtemperature and pressure for these penetrations.

Per Attachment 2 of specification X4AQ1O,the emergency operation design containment pressure is 50 psig.The evaluation in Attachment F of this calc demonstrates that the pipe penetrations should notfail due to a containment pressure of 52 psig.VEGP Condition Report 876376 has been submitted to review and resolve this difference between the Containment and the piping penetration accident design pressure criteria.

Thereare no operability or functionality issues because the peak containment DBA pressure is -37psig (VEGP FSAR Tables 6.2.1-1 & 6.2.1-66).

Electrical Penetrations Per section 3.1.3 of DC-1818, the electrical penetration assemblies shall withstand the pressure, temperature, and environmental conditions resulting from a DBA without exceeding the electrical penetration design leakage rate.Per section E3.6.2 of specification X3AROI-E3, the electrical penetration design leakage rate is0. 01 cc/sec at DBA conditions.

CONTAINMENT CLOSURE no.t established.

Basis: NEI 99-01 Rev 6, page 81.

V12, Page 4 of17Southern Nuclear Design Calculation SPlant: Vogtle Unit: 1&2 ICalculation Number: X6CNA15 ISheet: F-IAttachment F -Evaluation of 52 psig Pressure on Mechanical Penetrations Introduction There is a discrepancy between the DBA Design Pressures for the Containment (52 psig persection 3.4.5 of DC-21 01) and the pipe penetrations (50 psig per Attachment 2 of specification X4AQ1 0).This attachment evaluates the effect of a 52 psig Containment pressure on the pipepenetrations.

Conclusions The compressive and shear loads imposed by a 52 psig Containment pressure on the Unit1&2 pipe penetrations' welds are well below their allowable loads, less than -4% and -30%respectively.

Thus, the pipe penetrations are expected to maintain containment integrity at 52psig.MethodA Type I pipe penetration is shown below:iQlrt & 4N[From 1X4DL4A014

& 2X4DL4A014]

The weakest point of the penetration sleeve is the weld between the penetration sleeve andthe containment liner. If the loads imposed by containment pressure on these welds are lessthan the weld strength, the penetration is expected to maintain containment integrity.

From page 443 of "Strength of Materials":

"The strength of a butt weld is equal to the allowable stress multiplied by the product of the length of the weld times the thickness of the thinnerplate of the joint. The American Welding Society specifies allowable stresses of 20,000 psi intension or compression and 13,600 psi in shear."The specifications for Containment liner welds are likely to be more stringent (i.e., higherallowable stresses) than the values in this textbook.

Using these textbook values isconservative for the purposes of this evaluation:

establishing an allowable limit.

V12, Page 5 of 17Southern Nuclear Design Calculation IPlant: vogtle unit: 1&2 1Calculat°n Nubr: X6CNAI5 sheet: F-2 IAttachment F -Evaluation of 52 psig Pressure on Mechanical Penetrations These allowable stresses are mostlikely specified at standard temperature (68 F or 20 C). The maximum fluidtemperature passing through one ofthese penetrations is 557 F (-290 F).Per VEGP FSAR Table 6.2.1-1, thepeak DBA containment temperature is250 F (-120 C). The yield strength ofsteel decreases with increasing temperature as shown in therepresentative graph to the right.Reducing the above allowable stressesby 15% conservatively addresses theeffect of increased temperature 1,11.00,90,8I-eI-U)0,750 200 400 600Temperature

°CVariation of ultimate strength (Su) and yield strength (Sy)with ratio of operalin temp/Iroom temp (ST/SmT)http:l/www.roymech.co.ukiUsefulTables/Matter/Temperature effects.h~tnl The circumferential weld length (Lw) is calculated as followsLw=ix IDwhereID = Inside diameter of penetration sleeve = OD -2 x tD = OD of penetration sleeve (inches)t = penetration wall thickness (inches)The weld compressive strength (Fc Ibf) is calculates as follows:Fc= [a;cnom X ftemp] X Lw X twhere0c-nom = Nominal allowable compressive stress (20,000 psi)ftemp = Reduction due to increased temperature

= 0.85 = 1 -0.15Lw= Weld length (inches)T = Weld thickness (inches)

= Wall thickness (inches)The weld shear strength (Fs Ibf) is calculates as follows:Fs = [O's-nom X ftemp] X Lw X twhere0s-nora = Nominal allowable shear stress (13,600 psi)fternp = Reduction due to increased temperature

= 0.85 = 1 -0.15Lw= Weld length (inches)T = Weld thickness (inches)

= Wall thickness (inches)

V12, Page 6 of 17Southern Nuclear Design Calculation Plnt Votl Unit: 1& CacltoIume:XCA5sheet:

F-Attachment F -Evaluation of 52 psig Pressure on Mechanical Penetrations The Containment pressure (Pctmt psig) exerts a compressive load (Pc lbf) on the end of thepenetration sleeve. Using the sleeve outside diameter (D in the above figure) maximizes thisload:Pc = Pctmt x H x D2/4The Containment pressure (Pctmt psig) exerts a shear load (Ps Ibf) along the length of thepenetration sleeve. Using the overall sleeve length (L in the above figure) maximizes this load:Ps = Pctmt x H- x D x LEvaluation The effect of a 52 psig Containment pressure on the Unit 1 and Unit 2 pipe penetrations arecalculated in Excel spreadsheets Attachment F1 and Attachment F2.References F1. IX4DL4A0I3, Revision 7, "Containment Building Unit I Containment Wall PipePenetration Design List"F2. IX4DL4A014, Revision 9, "Containment Building Unit I Containment Wall PipePenetration Design List"F3. 2X4DL4A013, Revision 5, "Containment Building Unit 2 Containment Wall PipePenetration Design List"F4. 2X4DL4A014, Revision 4, "Containment Building Unit 2 Containment Wall PipePenetration Design List"F5. Singer, "Strength of Materials,"

second edition, 1962 V12, Page 7 of 17X6CNAI5 ATTACHMENT F SHEET F-4Bornt, ButchFrom: Jani, Yogendra M.Sent: Tuesday, October 14, 2014 4:52 PMTo: Borer, ButchCc: Patel, V. R.; Evans, William P. (SNC Corporate);

Lambert, David Leslie

Subject:

FW: VEGP Pipe Penetration EvalButch,i concur with your methodology used to evaluate 52 psig pressure on MeChanical Penetrations depicted on drawings 1X4DL4AO14

& 2X4DL4A014.

The loads imposed on the weakest point (weld)of penetrations are less than the weld strength.

The penetrations shall exceed the requirements ofASME Section I1I code. So the penetrations are in compliance with specification no. X4AQ10.Therefore, I agree with your conclusion that the pipe penetrations are expected to maintain structural integrity of containment integrity at 52 psig.Thank you,Vogeodra JaniSNCFleet Des -Safety Anl & Mech205.992.5125 office205.410.9806 mobile V12, Page 8 of 17SHEET Fl-iX6CNAI15ATTACHMENT F1Evaluation of 52 psig Pressure on UI Pipe Penetrations L = Overall Length of Penetration Sleeve (inches)D = Penetration Outside Diameter (inches)t = Penetration Sleeve Wall Thickness (inches)ID = Penetration Inside Diameter (inches)ID= D- (2 xt)Lw= Weld Length (inches)Lw= ix IDac = Allowable compressive stress (psi)a3c = a3c-nom X ftemp

= 20,000 psi = Nominal allowable comprssive stressftemp = 0.85 = Reduction due to increased temperature ac = 17,000 psia's =a's =Allowable shear stress (psi)O's-fara X ftemp

= 13,600 psi = Nominal allowable comprssive stressftemp = 0.85 = Reduction due to increased temperature 11,560 psiFc = O'c x Lw x t = Allowable Compressive Load (Ibf)Fs = as x Lw x t = Allowable Shear Load (Ibf)Pctmnt =52 psig = Containment PressureP0 = Compressive Load (lbf)Pc = Pctint x LI x [(D^2)/4]

Ps = Shear Load (Ibf)Ps = Pctmt X (LI X D) X L V12, Page 9 of 17SHEET F1-2X6CNAI5ATTACHMENT F1Evaluation of 52 psig Pressure on U1 Pipe Penetrations Compression LI I= Tp V orVII peerto;dimension B used instead of LType V Penetration:

Per IX4DL4A0 14, there is no dimension L; use B insteadType VII Penetration:

If dimension L not provided on 1X4DL4A013 or IX4DL4A014, use BinsteadPEN # Type L D t ID Lw Fc Pc PclFc1-4 I 48.25 56.000 1.500 53.000 167 4.2E+06 1.3E+05 0.0305 VII 8.250 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.0247-10 I 18.000 18.000 0.500 17.000 53 4.5E+05 1.3E+04 0.02911& 12 III 11.750 12.750 0.375 12.000 38 2.4E+05 6.6E+03 0.02813 III 12.750 12.750 0.375 12.000 38 2.4E+05 6.6E+03 0.02814 VII 15.250 10.750 0.365 10.020 31 2.OE+05 4.7E+03 0.02415 II 15.750 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.02416 -17 V 8.250 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.02418-21 I 37.500 34.000 1.500 31.000 97 2.5E+06 4.7E+04 0.01922 II 15.750 10.750 .0.365 10.020 31 2.0E+05 4.7E+03 0.02423 I 14.930 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.02424 II 15.750 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.02425 V 8.250 18.000 0.500 17.000 53 4.5E+05 1.3E+04 0.02928 &29 II 16.750 18.000 0.500 17.000 53 4.5E+05 1.3E+04 0.02930 & 31 II 18.750 14.000 0.438 13.124 41 3.1E+05 8.0E+03 0.02632 II 15.750 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.02433 II 18.750 14.000 0.438 13.124 41 3.1E+05 8.0E+03 0.02634 & 35 II 19.250 20.000 0.500 19.000 60 5.1E+05 1.6E+04 0.03240 II 16.750 18.000 0.500 17.000 53 4.5E+05 1.3E+04 0.02941 II 15.750 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.02442 II 12.000 10.750 0.365 10.020 31 2.OE+05 4.7E+03 0.02443-46 II 18.750 18.000 0.500 17.000 53 4.5E+05 1.3E+04 0.029V 8.250 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.02448 I 15.750 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.02449 II 15.750 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.02450 I 15.750 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.02451 -55 I 15.750 10.750 0.365 10.020 31 2.OE+05 4.7E+03 0.024 V12, Page 10 of 17SHEET F1-3X6CNA15ATTACHMENT F1Evaluation of 52 psig Pressure on U1 Pipe Penetrations Compression

[111I= TyeV orViipntain dimension B used instead of LType V Penetration:

Per 1X4DL4AO014, there is no dimension L; use B insteadType VII Penetration:

if dimension L not provided on 1X4DL4A013 or 1X4DL4AO14, use BinsteadPEN # Type JLJD~ t IDJ__ Fc_ Pc JPcIFc56 I__44.750134.000I1.500 31.0001 97 2.5E+06 4.7E+04 0.01957 & 58 I 32.000 24.000 1.000 22.000 69 1 .2E+06 2.4E+04 0.02059 & 60 I 28.000 26.000 1.000 24.000 75 1 .3E+06 2.8E+I04 0.02261 V 8.250 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.02462 &63 II 15.750 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.02464 VI 13.250 10.750 0.365 10.020 31 2.0E+05 4.7E+i03 0.02466 V j8.250 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.02467 III 12.750 12.750 0.375 12.000 38 2.4E+05 6.6E+03 0.02868 II 16.000 18.000 0.500 17.000 53 4.5E+05 1.3E+04 0.02969 -73 III 12.750 12.750 0.375 12.000 38 2.4E+05 6.6E+03 0.02875 V 8.250 10.750 0.365 10.020 131 2.0E+'05 4.7E+03 0.02476 V 8.250 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.02477 & 78 II 15.750 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.02479 II 12.000 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.02480 II 16.000 18.000 0.500 17.000 53 4.5E+05 1.3E+'04 0.02981 II 12.000 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.02482 V 8.250 18.000 0.500 17.000 53 4.5E+05 1.3E+04 0.02983 & 84 10.000 24.000 0.500 23.000 72 6.1E+05 2.4E+04 0.03885 III 12.750 12.750 0.375 12.000 38 2.4E+05 6.6E+03 0.02886 III 23.670 12.750 0.375 12.000 38 2.4E+05 6.6E+03 0.02887 II 16.000 18.000 0.500 17.000 53 4.5E+05 1.3E+04 0.02988 VII 15.250 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.02490 VII 8.250 10.750 0.365 10.020 31 2.0E+05 4.7E+'03 0.02491 -98 II 19.750 18.000 0.750 16.500 52 6.6E+05 1.3E+04 0.020100 9.250 4.500 0.237 4.026 13 5.1E+04 8.3E+02 0.016101 -104 I 20.550 18.000 0.750 16.500 52 6.6E+05 1.3E+04 0.020Maximum PclFc = 0.038 V12, Page 11 of 17SHEET FI-4X6CNAI15ATTACHMENT F1Evaluation of 52 psig Pressure on UI Pipe Penetrations Shear[111= Type V or VII penetration; dimension B used instead of LType V Penetration:

Per IX4DL4A0 14, there is no dimension L; use B insteadType VII Penetration:

If dimension L not provided on 1X4DL4A013 or 1X4DL4A014, use BinsteadPEN # IType L 0 t ID Lw Fs Ps JPs/Fs1 -4 I 48.250 56.000 1.5001 53.0001167 2.9E+06 4.4E+05 0.1535 VII 8.250 10.750 0.365 10.020 31.5 1.3E+05 1.4E+04 0.1097- 10 I 18.000 18.000 0.500 17.000 53.4 3.1E+05 5.3E+04 0.17111& 12 III 11.750 12.750 0.375 12.000 37.7 1.6E+05 2.4E+04 0.15013 III 12.750 12.750 0.375 12.000 37.7 1.6E+05 2.7E+04 0.16314 VII 15.250 10.750 0.365 10.020 31.5 1.3E+05 2.7E+04 0.20215 II 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.20816- 17 V 8.250 10.750 0.365 10.020 31.5 1.3E+05 1.4E+04 0.10918-21 I 37.500 34.000 1.500 31.000 97.4 1.7E+06 2.1E+05 0.12322 II 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.20823 I 14.930 10.750 0.365 10.020 31.5 1.3E+05 2.6E+04 0.19724 II 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.20825 V 8.250 18.000 0.500 17.000 53.4 3.1E+05 2.4E+04 0.07928 & 29 II 16.750 18.000 0.500 17.000 53.4 3.1E+05 4.9E+04 0.16030 & 31 II 18.750 14.000 0.438 13.124 41.2 2.1E+05 4.3E+04 0.20532 II 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.20833 II 18.750 14.000 0.438 13.124 41.2 2.1E+05 4.3E+04 0.20534 & 35 II 19.250 20.000 0.500 19.000 59.7 3.5E+05 6.3E+04 0.18240 II 16.750 18.000 0.500 17.000 53.4 3.1E+05 4.9E+04 0.16041 II 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.20842 II 12.000 10.750 0.365 10.020 31.5 1.3E+05 2.1E+04 0.15943-46 II 18.750 18.000 0.500 17.000 53.4 3.1E+05 5.5E+04 0.17947 V 8.250 10.750 0.365 10.020 31.5 1.3E+05 1.4E+04 0.10948 I 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.20849 II 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.20850 I 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.20851 -55 I 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.208 V12, Page 12 of 17SHEET Fl-5X6CNA1 5ATTACHMENT FlEvaluation of 52 psig Pressure on UI Pipe Penetrations ShearI ... I=Type V or VII penetration; dimension B used instead of LType V Penetration:

Per IX4DL4A0 14, there is no dimension L; use B insteadType VII Penetration:

If dimension L not provided on IX4DL4A013 or 1X4DL4A014, use BinsteadPEN # Type L D t ID Lw Fs Ps P5/F556 I 44.750 34.000 1.500 31.000 97.4 1.7E+06 2.5E+05 0.14757 & 58 I 32.000 24.000 1.000 22.000 69.1 8.0E+05 1.3E+05 0.15759 & 60 I 28.000 26.000 1.000 24.000 75.4 8.7E+05 1.2E+05 0.13661 V o8.25'0 10.750 0.365 10.020 31.5 1.3E+05 1.4E+04 0.10962 &63 II 15.750 10.750 0.365 10.020 31.5 i.3E+05 2.8E+04 0.20864 VI 13.250 10.750 0.365 10.020 31.5 1.3E+05 2.35+04 0.175,,66 ...IV 8.250; 10.750 0.365 10.020 31.5 1.3E+05 1.4E+04 0.10967 III 12.750 12.750 0.375 12.000 37.7 1.6E+05 2.7E+04 0.16368 II 16.000 18.000 0.500 17.000 53.4 3.1E+05 4.7E+04 0.15269 -73 III 12.750 12.750 0.375 12.000 37.7 1.6E+05 2.7E+04 0.163:75: V °8,250 10.750 0.365 10.020 31.5 1.3E+05 1.4E+04 0.109:V 8.250 10.750 0.365 10.020 31.5 1.3E+05 1.4E+04 0.10977 & 78 II 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.20879 II 12.000 10.750 0.365 10.020 31.5 1.3E+05 2.IE+04 0.15980 II 16.000 18.000 0.500 17.000 53.4 3.IE+05 4.7E+04 0.15281 II 12.000 10.750 0.365 10.020 31.5 1.3E+05 2.1E+04 0.159182 ....8.250 18.000 0.500 17.000 53.4 3.1E+05 2.4E+04 0.07983 & 84 10.000 24.000 0.500 23.000 72.3 4.2E+05 3.9E+04 0.09485 III 12.750 12.750 0.375 12.000 37.7 1.6E+05 2.7E+04 0.16386 III 23.670 12.750 0.375 12.000 37.7 1.6E+05 4.9E+-04 0.30287 II 16.000 18.000 0.500 17.000 53.4 3.IE+05 4.7E+04 0.15288: VII i,.250' 10.750 0.365 10.020 31.5 1.3E+05 2.7E+04 0.202'90 " VII 8.250 10.750 0.365 10.020 31.5 1.3E+05 1.4E+04 0.10991 -98 II 19.750 18.000 0.750 16.500 51.8 4.5E+05 5.8E+04 0.129100 9.250 4.500 0.237 4.026 12.6 3.5E+04 6.8E+03 0.196101,-104 I 20.550 18.000 0.750 16.500 51.8 4.5E+05 6.0E+04 0.134Maximum PslFs = 0.302 V12, Page 13 of 17X6CNA15 ATTACHMENT F2 SHEET F2-1Evaluation of 52 psig Pressure on U2 Pipe Penetrations L = Overall Length of Penetration Sleeve (inches)D = Penetration Outside Diameter (inches)t = Penetration Sleeve Wall Thickness (inches)ID = Penetration Inside Diameter (inches)I D= D-(2xt)Lw= Weld Length (inches)Lw= fixiD-c= Allowable compressive stress (psi)c'c = O'c-nom X ftempac-norn = 20,000 psi = Nominal allowable comprssive stressftermp = 0.85 = Reduction due to increased temPerature O-c = 17,000 psi = Allowable compressive stressas = Allowable shear stress (psi)a's = a's-nom X ftempa'c-nomn

= ,I13,600 psi = Nominal allowable comprssive stressftemp = 0.85 = RedUction due to increased temperature O's = 11,560 psi = Allowable shear stressFc = ac x Lw x t = Allowable Compressive Load (Ibf)Fs = as x Lw x t = Allowable Shear Load (Ibf)Pctmnt = 52 psig = Containment PressurePc = Compressive Load (lbf)Pc = Pctmt x H x [(D^2)/4]

Ps = Shear Load (Ibf)Ps = Pctmt X (H- X D) X L V12, Page 14 of 17SHEET F2-2X6CNA15SATTACHMENT F2Evaluation of 52 psig Pressure on U2 Pipe Penetrations Compression I I= Type V or VII penetration; dimension B used instead of LiType V Penetration:

Per 2X4DL4A0 14, there is no dimension L; use B insteadType VII Penetration:

If dimension L not provided on 2X4DL4A013 or 2X4DL4A014, use BinsteadPEN # Type L D t ID Lw FcPc PclFc1 -4 I 48.25 56.000 1.500 53.000 167 4,2E+06 1.3E+05 0.0305 VII 8.250 10.750 0.365 10.020 31 2,0E+05 4,7E+03 0.0247- 10 I 18.000 18.000 0.500 17.000 53 4.5E+05 1,3E+04 0.02911& 12 III 11.750 12.750 0.375 12.000 38 2,4E+05 6,6E+03 0.02813 III 12.750 12.750 0.375 12.000 38 2.4E+05 6.6E+03 0.02814 VII 15.250 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.02415 II 15.750 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.02416 -17 V 8.250 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.02418-21 I 37.500 34.000 1.500 31.000 97 2.5E+06 4,7E+04 0.01922 II 15.750 10.750 0.365 10.020 31 2,0E+05 4,7E+03 0.02423 I 14.930 10.750 0.365 10.020 31 2.0E+05 4,7E+03 0.02424 II 15.750 10.750 0.365 10.020 31 2,0E+05 4,7E+03 0.02425 ,V 8.250 18.000 0.500 17.000 53 4.5E+05 1,3E+04 0.02928 & 29 II 16.750 18.000 0.500 17.000 53 4.5E+05 1,3E+04 0.02930 & 31 II 18.750 14.000 0.438 13.124 41 3.1E+05 8,0E+03 0.02632 II 15.750 10.750 0.365 10.020 31 2,0E+05 4,7E+03 0.02433 II 18.750 14.000 0.438 13.124 41 3,1E+05 8,0E+03 0.02634 & 35 II 19.250 20.000 0.500 19.000 60 5.IE+05 1,6E+04 0.03240 II 16.750 18.000 0.500 17.000 53 4,5E+05 1,3E+04 0.02941 II 15.750 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.02442 II 12.000 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.02443 -46 II 18.750 18.000 0.500 17.000 53 4.5E+05 1.3E+04 0.029"47 V 8.250 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.02448 I 15.750 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.02449 II 15.750 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.02450 I 15.750 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.02451 -55 I 15.750 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.024 V12, Page 15 of 17SHEET F2-3X6CNA15SATTACHMENT F2Evaluation of 52 psig Pressure on U2 Pipe Penetrations Compression

~=TpeorI=

pentraIon dimension Bused instead of LType V Penetration:

Per 2X4DL4A0 14, there is no dimension L; use B insteadType VII Penetration:

If dimension L not provided on 2X4DL4A013 or 2X4DL4A014, use BinsteadPEN # Type L D t ID LFcPc Pc/Fc56 I 44.750 34.000 1.500 31.000 97 2.5E+06 4.7E+04 0.01957 & 58 I 32.000 24.000 1.000 22.000 69 1.2E+06 2.4E+04 0.02059 & 60 1 28.000 26.000 1.000 24.000 75 1 .3E+06 2.8E+04 0.02261 V 8.250 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.02462 &63 II 15.750 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.02464 VI 13.250 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.02466 V 8.250 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0,02467 III 12,750 12,750 0.375 12,000 38 2,4E+05 6.6E+03 0,02868 II 16.000 18.000 0.500 17.000 53 4.5E+05 1.3E+04 0.02969-73 III 12.750 12.750 0.375 12.000 38 2.4E+05 6.6E+03 0.02875 V 8.250 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.02476 V 8.250 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.02477 &78 II 15.750 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.02479 II 12,000 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.02480 II 16.000 18.000 0.500 17.000 53 4.5E+05 1.3E+04 0.02981 II 12.000 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.02482 V 8.250 18.000 0.500 17.000 53 4.5E+05 1.3E+04 0.02983 &84 10.000 24.000 0.500 23.000 72 6.IE+05 2.4E+04 0.03885 III 12.750 12.750 0.375 12.000 38 2.4E+05 6.6E+03 0.02886 III 23.670 12.750 0.375 12.000 38 2.4E+05 6.6E+03 0.02887 II 16.000 18.000 0.500 17.000 53 4.5E+05 1.3E+04 0.02988 VII 15.250 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.02490 VII 8.250 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.02491 -98 II 19.750 18.000 0.750 16.500 52 6.6E+05 1.3E+04 0.020100 9.250 4.500 0.237 4.026 13 5.1E+04 8.3E+02 0.016101 -104 I 20.550 18.000 0.750 16.500 52 6.6E+05 1.3E+04 0.020Maximum PdIFc = 0.038 V12, Page 16 of 17SHEET F2-4X6CNA15SATTACHMENT F2Evaluation of 52 psig Pressure on U2 Pipe Penetrations ShearType V orVillpntain dimension B used instead of LType V Penetration:

Per 2X4DL4A014, there is no dimension L; use B insteadType VII Penetration:

If dimension L not provided on 2X4DL4A013 or 2X4DL4A014, use BinsteadPEN # Type L D t ID Lw Fs Ps PslFs1 -4 I 48.250 56.000 1.500 53.000 167 2.9E+06 4.4E+05 0.1535 VII 8.250 10.750 0.365 10.020 31.5 1,3E+05 1.4E+04 0.1097-10 I 18.000 18.000 0.500 17.000 53.4 3.1E+05 5.3E+04 0.17111& 12 III 11.750 12.750 0.375 12.000 37.7 1.6E+05 2.4E+04 0.15013 III 12.750 12.750 0.375 12.000 37.7 1.6E+05 2.7E+04 0.16314 VII 15.250 10.750 0.365 10.020 31.5 1.3E+05 2.7E+04 0.20215 II 15.750 10.750 !0.365 10.020 31.5 1.3E+05 2.8E+04 0.20816-17 V 8.250 10.750 0.365 10.020 31.5 1.3E+05 1.4E+04 0.10918-21 1 37.500 34.000 1.500 31.000 97.4 1.7E+06 2.IE+05 0.12322 II 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.20823 I 14.930 10.750 0.365 10.020 31.5 1.3E+05 2.6E+04 0.19724 II 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.20825 V 8.250 18.000 0.500 17.000 53.4 3.1E+05 2.4E+04 0.07928 &29 II 16.750 18.000 0.500 17.000 53.4 3.1E+05 4.9E+04 0.16030 & 31 II 18.750 14.000 0.438 13.124 41.2 2.1E+05 4.3E+04 0.20532 II 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.20833 II 18.750 14.000 0.438 13.124 41.2 2.IE+05 4.3E+04 0.20534 & 35 II 19.250 20.000 0.500 19.000 59.7 3.5E+05 6.3E+04 0.18240 II 16.750 18.000 0.500 17.000 53.4 3.IE+05 4.9E+04 0.16041 II. 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.20842 II 12.000 10.750 0.365 10.020 31.5 1.3E+05 2.1E+04 0.15943-46 II 18.750 18.000 0.500 17.000 53.4 3.1E+05 5.5E+04 0.17947 V 8.250 10.750 0.365 10.020 31.5 1.3E+05 1.4E+04 0.10948 I 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.20849 II 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.20850 I 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.20851-55 I 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.208 V12, Page 17 of 17SHEET F2-5X6CNA1 5ATTACHMENT F2Evaluation of 52 psi9 Pressure on U2 Pipe Penetrations Shear[111= Type V or VII penetration; dimension B used instead of LType V Penetration:

Per 2X4DL4A014, there is no dimension L; use B insteadType VII Penetration:

If dimension L not provided on 2X4DL4A013 or 2X4DL4A014, use BinsteadPEN # Type L D t ID L FsPs PslFs56 I 44.750 34.000 1.500 31.000 97.4 1.7E+06 2.5E+05 0.14757 & 58 I 32.000 24.000 1.000 22.000 69.1 8.0E+05 1.3E+05 0,15759 & 60 I 28.000 26.000 1.000 24.000 75.4 8.7E+05 1.2E+05 0.13661 V 8.250 10.750 0.365 10.020 31.5 1.3E+05 1.4E+04 0.10962 &63 II 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.20864 VI 13.250 10.750 0.365 10.020 31.5 1.3E+05 2.3E+04 0.17566 V 8.250 10.750 0.365 10.020 31.5 1.3E+05 1.4E+04 0.10967 III 12.750 12.750 0.375 12.000 37.7 1.6E+05 2.7E+04 0.16368 II 16.000 18.000 0.500 17.000 53.4 3.1E+05 4.7E+04 0.15269-73 III 12.750 12.750 0,375 12.000 37.7 1.6E+05 2.7E+04 0.16375 V 8.250 10.750 0.365 10.020 31.5 1.3E+05 1.4E+04 0.10976 V 8.250, 10.750 0.365 10.020 31.5 1.3E+05 1.4E+04 0.10977 &78 II 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.20879 II 12.000 10.750 0.365 10.020 31.5 1.3E+05 2.1E+04 0.15980 II 16.000 18.000 0.500 17.000 53.4 3.1E+05 4.7E+04 0.15281 II 12.000 10.750 0.365 10.020 31.5 1.3E+05 2.1E+04 0.15982 V 8.250 18.000 0.500 17.000 53.4 3.1E+05 2.4E+04 0.07983 & 84 10.000 24.000 0.500 23.000 72.3 4.2E+05 3.9E+04 0.09485 III 12.750 12.750 0.375 12.000 37.7 1.6E+05 2.7E+04 0,16386 III 23.670 12.750 0.375 12.000 37.7 1.6E=+05 4.9E=+04 0.30287 II 16.000 18.000 0.500 17.000 53.4 3.1E+05 4.7E+04 0.15288 VII 15.250 10.750 0.365 10.020 31.5 1.3E=+05 2.7E+04 0.202,90, VII 8.250 10.750 0.365 10.020 31.5 1.3E+05 1.4E+04 0.10991 -98 II 19.750 18.000 0.750 16.500 51.8 4.5E+05 5.8E+04 0.129100 9.250 4.500 0.237 4.026 12.6 3.5E+04 6.8E+03 0.196101 -104 I 20.550 18.000 0.750 16.500 51.8 4.5E+05 6.0E+-04 0.134Maximum PslFs 0.302 I-+/- 4.4.4 .J~XH~_ ~ Th.7r~.jVi3Page 1 of 1T<-H........

.....--I ALVI W. VOCTHE P1AJhC.;a i i a i V1 4Page 1 of 2provided to prevent explosive concentrations of hydrogen during battery charging.

The ventilation system shall be adequate to maintain the hydrogen concentration below 2 % in accordance with IEEE 484.*Each room shall be provided with a shower and an eye-wash basin in case personnel come in contact with battery acid.*To prevent possible hydrogen explosion, switches and receptacles shall not be locatedinside battery rooms.3.2 POWER GENERATION DESIGN BASESNone3.3 MAJOR COMPONENT DESIGN BASES3.3.1 The Class IE dc 125-V dc system and all equipment are classified as Seismic Category 1and Safety Class 1E. The system equipment shall be capable of continuous operation between 100 and 140 V dc, except vital ac bus inverters, the reactor trip switchgear

control, residual heat removal (RHR) isolation valve inverters, and the turbine-driven auxiliary feedwater pump control may be allowed to operate over a dc input voltagerange of 105 to 14OV dc.The 25 kVA inverters provide power to RIHR isolation valves. The breakers for thesevalves are located in "Trip" position during normal plant operation; they do not operateduring the duty cycle time of the battery, or for any LOCA/LOOP or SBO copingscenario.

3.3.2 Reqiuirements for Batteries 3.3.2.1 Battery size shall be determined in accordance with the method indicated in IEEE 485.3.3.2.2 Each Class 1E battery shall have sufficient capacity to independently supply the requiredloads for a loss-of-coolant

accident, loss of offsite power, or main steam line break for aduration of 2.75 hr. Each Class 1E battery shall have sufficient capacity to independently supply the required loads for a station blackout for a duration of 4 hr.3.3.2.3 Initial battery capacity shall be 25 % greater than required according to the calculation method indicated in IEEE 485 to allow for aging and extend the time interval for batteryreplacement as required by the battery replacement criteria of IEEE 450.3.3.2.4 Batteries shall be sized to provide their required output at 70°F.3.3.2.5 A margin of 10 % load growth shall be initially included in the sizing of each battery.DC-i1806 6VR16VER 13 V14Page 2 of 23.3.3.7 A dc ammeter, dc voltmeter, dc overvoltage relay, ac power "on" light, and acundervoltage relay shall be provided on each charger.

The relays shall alarm in thecontrol room.3.3.3.8 The chargers shall be suitable for parallel operation so that each subsystem's redundant charger can be manually put into operation in conjunction with the normal charger torecharge a discharged battery in a shorter amount of time (if allowed by themanufacturer).

3.3.3.9 The battery charger shall prevent the charger from becoming a load on the battery due toa power feedback during loss of ac power to the chargers.

3.3.3.10 ac breakers shall be provided to protect the charger from internal faults and to isolate thecharger from the ac source.3.3.3.11 dc breakers shall be provided to protect the battery and charger from internal chargerfaults and to isolate the charger from the dc system.3.3.4 Requirements for 125-V dc Metal-Enclosed Switchgear 3.3.4.1 There is one dc switchgear lineup for each subsystem.

It shall be connected to the battery,the normal battery charger, and the redundant battery charger associated with that train.The switchgear shall feed MCCs, dc distribution panels, and the inverter for the vitalinstrumentation system (DC- 1807).3.3.4.2 The switchgear air circuit breakers shall be equipped with direct-acting, dual-magnetic, overcurrent tripping devices providing adjustable overcurrent and short-circuit protection.

3.3.4.3 The 125-V dc switchgear breakers shall serve as a means for energizing and deenergizing power sources and loads connected to the 125-V dc switchgear bus. The switchgear shallalso provide suitable protection for the loads during overload and short-circuit conditions.

3.3.4.4 Trains C and D shall each provide 125-V dc power to an associated 480-V, 3-phaseinverter for RHR isolation valves.3.3.4.5 The vital ac bus inverters may be allowed to operate over a dc input voltage range of 105to 140 V dc. The dc feeder cables shall be designed to maintain a minimum of 105 V dcduring the entire battery load profile.3.3.4.6 The RHR isolation valve inverters may be allowed to operate over a range of 10._5 to140 V dc. The dc feeder cables to the RHR isolation valve inverters shall be designed tomaintain a minimum of 105 V dc at the RH-R isolation valve inverters.

3.3.5 125-V dc MCCs3.3.5.1 One MCC shall be provided for each train A, B, and C subsystem.

It shall be connected to the switchgear associated with that subsystem.

The MCC shall feed motor-operated DC-1806 8VR18VER 13 X6CNA1 5Attachment L

C2I.l~at~inn fnr I:W.I I1SHEET L-6 v15Page 1 of 3CALC NO. SNC024-CALC-004 VEGP DETERMINATION OFEMERGENCY ACTION LEVEL REV. 0I0 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 6.2-3] are provided below in Table 5-1. Thesesource values are scaled to develop the emergency action levels for initiating condition E-HU1.Table 5-1 Technical Specification (Neutron

+Gamma) Dose Rate Limits for HI-STORM100 and HI-TRAC 125Number of Technical Specification LoatonjMeasurements j Limit (mrem/hr)

HI-TRAC 1.25 __________

Side -Mid -height 4 472.7Top 1 4 j102.4HI-STORM 100 ____ _____Side -60 inches below mid-height 4 87Side -Mid -height 4 88.9Side -60 inches above mid-height 4 54.8Top -Center of lid 1 24.5Top -Radially centered 4 29.2Inlet duct 4 178.8Outlet duct 4 64.5 X6CNA1 5Attachment LENERCON Calculation for E-HU1SHEET L-7 v15Page 2 of 3CALC NO. SNC024-CALC-004 VEGP DETERMINATION OF ____________

EMERGENCY ACTION LEVEL REV. 00 E N E R C 0 N FOR INITIATING CONDITION E- -__________

PAGE NO. Page 7of 86.0 Methodology The "on-contact" dose rates from the technical specification for the HI STORM-i100 casksystem are scaled by a factor of 2, as specified in NEI 99-01 Rev. 6 [1], for use in initiating condition E-HUI.

X6CNA1 5Attachment LFNFRf.ON

('nlrmiitinn fnr F-HI- 11SHEET L-8V1Page 3 of 3CALC NO. SNC024-CALC-004 VEGP DETERMINATION OF -____________

EMERGENCY ACTION LEVEL REV. 00 E N E R C 0 N FOR INITIATING CONDITION E-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 LimitsTechnical LoainSpecification Scaling Calculated Value EALLoainLimit Factor (mrem/hr)

(mrem/hr)

_________________________

(mrem/hr) j____ _______________

HI-TRAC 125___ _____Side -Mid -height [ 472.7 2 1 945.4 [ 950Top 102.4 2 j 204.8 [ 200HI-STORM 100Side -60 inches below mid-height 87 2 174 170Side -Mid -height 88.9 2 177.8 180Side -60 inches above mid-height 54.8 2 109.6 110Top -Center of lid 24.5 2 49 50Top -Radially centered 29.2 2 58.4 60Inlet duct 178.8 2 357.6 360Outlet duct 64.5 2 129 1308.0 Computer SoftwareMicrosoft WORD 2013 is used in this calculation for basic multiplication.

V1 6Page 1 of 4Approved ByJ. B. StanleyEffective Date7/25/12~Voqgtle Electric Generating F-0 CRITICAL SAFETY FUNCTION SiF- 0.2CORE COOLINGPatProcedure versionPat19200-C 24.2Page Number'ATUS TREES5olSheet 1 of 1-4 GO TO11221-V~GO TO19221-CVLI5 FULL NOGE GREATERtHAN 41% ,E8S* .GO TO19222-C11.FI GO TO19222Z-CtVUI FULL NiGE GREATERTHAN 41% YE.GO TO* 19223-V~GO TOrYNAMIC Ha 9,Z-INGE Nho -4 RCPh-3 RCPh-2RCP E-1IRCF* 19223-CGI CSA TvPrinted February 2, 2016 at 16:09 Vi16Page 2 of 4Approved By Procedure VesoJ. B. Stanley Vogtle Electric Generating P..lantng 19200-C Veso24.2Effective Date Page Number7/25/12 F-0 CRITICAL SAFETY FUNCTION STATUS TREES6of1 Sheet 1 of 1F- 0.3HEAT SINKIKTOTAL AVAILABLE FEEDWATER FLOWTO SGs GREATERiTHAN 570 GPMNOYESNARROW RANGELEVEL IN AT LEASTrONE SG GREATERTHAN 10% (32%)GO TO1 9231 -CGO TO19232-CGO TO1 9233-C NOPRESSURE IN ALL SGsLESS THAN 1240 PSIG YELESS THAN 82% YESNOPRESSURE IN ALL SGsLESS THAN 1180 PSIGYESGO TO1 9234-CGO TO19235-CNARROW RANGE NOLEVEL IN ALL SGsGREATER THAN10% (32%) YESLSATPrinted February 2, 2016 at 16:09 V1 6Page 3 of 4Approved By Procedure VersionJ. B. Stanley Vogtle Electric Generating Plant 19200-C 24.2Effective Date Page Number7/25/12 F-0 CRITICAL SAFETY FUNCTION STATUS TREES7of1 Sheet 1 of 1F- 0.4INTEGRITY GO TO19241-C---- .) GO TOI 19241-CALL RCS WR COLD LEGTEMPERATURES GREATER THAN 2950F~QGO TO19242-C~*CSF SATTEMPERATURE DECREASE IN ALL NORCS COLD LEGS LESS1000F IN THEILAST iYES!60 MINUTESJGOTOI,11 19241-CALL RCS WR COLD LEG NOTEMPERATURES GREATERTHAN 265°F YESRCS PRESSURE LESS NOTHAN COLDOVERPRESSURE LIMIT 465 PSIG YES,IRCS WR COLD LEG NOTEMPERATURE GREATER THAN 2200F YESLL* (I'\ GO TO" "" 19242-CO CSF SAT-*CSF SATPrinted February 2, 2016 at 16:09 Vi16Page 4 of 4Approved By .....Procedure VersionJ. B. StanI~yVogtle Electric Generating Plant '19200-C 2.Effe tiv Datan ey P... ..age... :.. .... .... ... ...... NumbeEfeciv DteF-0 CRITICAL SAFETY FUNCTION STATUS TREES 9ag ofmb1r7/25/129of1 Sheet 1 of 1F- 0.5CONTAINMENT GO TO19251-C* )GO TOi 19251-CN.E N oPGO TOp 19252-CICONTAINMENT BUMP LEVEL LESSTHAN 196 iNCHESGO TO.......19253.CIbCSF SATPrinted February 2, 2016 at 16:09 V1 7Page 1 of 3Approved By p, 4Pocdr VersionW. L. Burmeister

... Vogtle Electric Generating Plant IS55039-c~rcdr 3.2Effective Date P~age Number0510712013 ISEISMIC MONITORING INSTRUMENTATION SYSTEM I 6 of 94.2 NORMAL OPERATION NONE4.3 NON-PERIODIC OPERATIONS NOTEThis subsection shall be initiated by 50022-C "Seismic Event Plan" orSupervisor's direction.

14.3.1 Retrieving Seismic DataNOTESKey Number 1-OP3-10 is in C&T. LIa. Verify the Event alarm on the Condor Control Unit screen isRED. Elb. Obtain the event charts and graphs from each recorder.

El4.3.2 Retrieving Seismic Data At ETNA (River Intake)a. Hook up a laptop to the uplink cable. Elb. Power on computer, THEN click on the ALTUS Quick Talkicon. Elc. In the ALTUS Status window, verify the alarm has beentriggered.

Eld. Save the file to a floppy disk. El(1) Click on the EVTI folder in the ALTUS directory window El(2) Highlight the EVTI file associated with the recordedevent El(3) Click on the Retrieve File button El(4) Download the file to a floppy disk ElPrinted February 15, 2016 at 16:59 V17Page 2 of 3Seismic Event Plan 50022-CVOGTLE Version 14.0Unit C Page 7of 244.0 INSTRUCTION 4.1 IDENTIFICATION OF SEISMIC EVENT1. The following indicators are available for determining whether or not seismicevent has occurred:

a. The Event Alarm on the Condor Control Unit screen is RED.b. The National Earthquake Information Center, located in Denver,Colorado, Telephone (303) 273-8500, confirms that an earthquake hasoccurred; (Must be called when any of the above indications of anearthquake is received) and to initiate this procedure if earthquake isconfirmed.

2. IF an earthquake is sensibly detected by control room personnel, initiate thisprocedure.

3. IF AOP 18036-C, Seismic Event, is in effect, then initiate this procedure.

Printed February 15, 2016 at 17:02 V1 7Page 3 of 3Seismic Event 18036-CVOGTLE Version 11Unit C Page 4of 9PURPOSEThe purpose of this procedure is to provide operator response following a seismic event and toinitiate an engineering analysis to determine the severity of the event.SYMPTOMS* Actuation of seismic monitor alarm.* Actuation of seismic instrumentation.

• Effects of earthquake heard or felt.MAJOR ACTIONS* Evaluate effects of seismic event.* Determine if shutdown of the units is required.

Printed February 15, 2016 at 17:05 VI18Page 1 of 5Southern Nuclear Design Calculation IPlant: Vogte Unit: 1&2 Icalculation Number: X6CNAI4 Isheet:;46 Miscellaneous Design Inputs21. Iodine boiling point = 184 C = -363 F

Reference:

Page B-I, "C3RC Handbook of Chemistry

& Physics"22. Density of Refueling Cavity and Spent Fuel Pool Water @ 130 F = 61.55 Ibm/cu ft

Reference:

See Attachment C32.23. Density of C3VCS letdown flow = 0.99 g/cc (Attachment C2)

Reference:

The density is used to convert the letdown activity from p.(Ci/g to ltCi/cc, whichare the units used by the C3VCS letdown rad monitor RE-48000 (Design Input #1 &Attachment CS5). Based on at-power CVCS letdown parameters from the Unit 1 and 2 IPCs(Attachment C35), the average temperature and pressure at the radiation measurement location are 98.5 F and 385 psig.24. Average Decay Gamma Energies for RE-48000 principle isotopes (Attachment C38)I rIsotopeAverageGammaEnergy(MeV)

Reference:

Brookhaven National Laboratory NationalNuclear Data Center decay data(http://www.orau

.or qlptp/PTP%20Libraryllibrary/DOE/bnl/nu clidedata/table.htm)

Copies of web pages in Attachment C81-131 0.3821-132 2.201-133 0.607I-134 2.501-135 1.55Co-580.975 Co-60 2.51Cs-134 1.55Cs-136 2.12Cs-i137 0.565Cs-1382.31Cs-138 2.31 Vi18Page 2 of 5Southern Nuclear Design Calculation SPlant: Vogtle Unit: 1&2 ICalculation Number: X6CNA14 ISheet: 61Recognition Category S: System Malfunctions Notice of Unusual EventSU4: Fuel Clad Degradation.

Operating Mode Applicability:

Power Operation (Mode 1)Startup (Mode 2)Hot Standby (Mode 3)Hot Shutdown (Mode 4)1 OR2Emergency Action Levels:SU4 EALI: CVCS Letdown radiation monitor RE-48000 reading greater than 5 pCi/ccindicating fuel clad degradation greater than Technical specification allowable limits.There are two Technical Specification limits on RCS coolant activity:

• SR 3.4.16.1:

Gross specific activity

< pCi/gm* SR 3.4.16.2:

Dose Equivalent 1-131 (DE 1-131) < 1.0 !iCi/gPer section B.3.4.16, page B3.4.16-2 of VEGP Tech Spec Bases, noble gasactivity in the reactor coolant assumes 1% failed fuel, which closely equalsthe LCO limit of 1 00/1s pCi/gm for gross specific activity.

The EAL threshold will be calculated for each Tech Spec limit condition.

Per pages 12 and 13of X6AZ01 A, theprinciple isotopesdetected by RE-48000 are 1-131, 1-133, Co-58, Co-60,Cs-134, and Cs-137.However, per SectionB-12-3-2 and FigureB-12-2 of 1X6AZ01-10004 & 2X6AZ01-10004, RE-48000 willdetect gammas ofenergies down to-0.1 MeV.St1 __ _ "__I-, -.. I _ _ _ _i c -i ..L =t 4 ..II; P.mIENKR4Y It ,VgL= t.VIFigure B-12-2Thus the other I, Co, and Cs isotopes listed in FSAR Table 11.1-2 should beincluded if their average decay gamma energies exceed 0.1 MeV.

V1 8Page 3 of 5Southern Nuclear Design Calculation iPlant: Vogtle U nit: 1&2 ICalculation Number: X6CNAI4 Sheet: 62Per LTR-CRA-06-179 attached to WEC-SNC letter GP-18006, the pre-MURPU coolant activities may be adjusted upward 2% to account for theincrease in core thermal power from 3565 MWt to 3636 MWt. Thus, the Coand Cs MURPU 1% defect activity are equal to their pre-MURPU 1% Defectactivities multiplied by 1.02.The Co and Cs activities corresponding to the 1.0 DE 1-131 TechSpec limit are the products of their MURPU 1% defect activities and theratio of the 1-131 DE 1-131 concentration to its equilibrium concentration (0.74/2.91).

The activities, expressed in j!iCi/g are summed and then multiplied by theCVCS letdown flow density (0.99 g/cc) to convert them to The EAL threshold is the minimum of the 1% Defect and the 1 .0 DE I-131 activities.

1.0 MURPU Pre-MURPU DE I-131 1% Defect 1% DefectIsotope Coolant Coolant CoolantActivity Activity ActivityI-131 0.74 2.91 ______I-132 0.75 2.96 ______I-133 1.41 5.561-134 0.18 0.69 ______I-135 0.69 2.72 ______Co-58 3.89E-03 1 .53E-02 1 .50E-02Co-60 4.93E-04 1 .94E-03 1 .90E-03Cs-134 5.97E-01 2.35 2.3Cs-I136 7.52E-01 2.96 2.9Cs-137 3.89E-01 1.53 1.5Total = 5.5 21.7 ptCi/gTotal = 5.5 21.5 i.LCi/ccCVCS Letdown Density =0.99g/ccSGiven the RG 1.97 R2 required system accuracy (Acceptance Criterion 3),the threshold is rounded down from 5.5 to 5 jltCi/cc.

NOTE: SU4 EAL2 not determined in this calculation.

V1 8Southern Nuclear Design Calculation Page 4 of 5SPlant: Vogtle Unit: 1&2 Calculation Number: X6CNA14 Sheet: C5-1Attachment C5 -VEGP 1&2 CVCS Letdown Radiation Monitor (RE-48000)

ReadingsU 1.... ... .aIIII-~' I~~-:-~a..

V1 8Southern Nuclear Design Calculation Page 5 of 5SPlant: Vogtle Unit: 1&2 Calculation Number: X6CNA14 Sheet: C5-2Attachment C5 -VEGP 1&2 CVCS Letdown Radiation Monitor (RE-48000)

Readings*]W11o=I~IMY Iin'~vu l~u~r ~.~tImP~.ii I~' ~'~IWE ~'~~jL 11U WOWW4m~ ~

V19Page 1 of 3RCS Specific Activity3.4.163.4 REACTOR COOLANT SYSTEM (RCS)3.4.16 RCS Specific ActivityLCO 3.4.16APPLICABILITY:

The specific activity of the reactor coolant shall be within limits.MODES 1 and 2,MODE 3 with RCS average temperature (Tavg) > 500°F.ACTIONS--------------------------

INlJLCO 3.0.4c is applicable.

I------------------

---CONDITION REQUIRED ACTION COMPLETION TIMEA. DOSE EQUIVALENT A.1 Verify DOSE Once per4 hoursI-131 > 1.0 p.Ci/gm.

EQUIVALENT I-131within the acceptable region of Figure 3.4.16-1.

ANDA.2 Restore DOSE 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />sEQUIVALENT I-131 towithin limit.B. Gross specific activity of B.1 Perform SR 3.4.16.2.

4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />sthe reactor coolant notwithin limit. AND8.2 Be in MODE 3 with 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />sTavg < 500°F.(continued)

Vogtle Units 1 and 23.4.16-1Amendment No. 137 (Unit 1)Amendment No. 116 (Unit 2)

V1 9Page 2 of 3RCS Specific Activity3.4.16ACTIONS (continued)

________________

__________

CONDITION REQUIRED ACTION COMPLETION TIMEC. Required Action and C.1 Be in MODE 3 with 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />sassociated Completion Tavg < 500°F.Time of Condition A notmet.O_.RDOSE EQUIVALENT 1-131 in theunacceptable region ofFigure 3.4.16-1.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.16.1 Verify reactor coolant gross specific In accordance withactivity_

100/I. !iCi/gm.

the Surveillance Frequency ControlProgramSR 3.4.16.2

---

--NOTE- --- --Only required to be performed in MODE 1.Verify reactor coolant DOSE EQUIVALENT I-131 In accordance withspecific activity

< 1.0 ,.tCi/gm, the Surveillance Frequency ControlProgramANDBetween 2 and6 hours after aTHERMAL POWERchange of _> 15% RTPwithin a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period(continued)

Vogtle Units 1 and 23.4.16-2Amendment No. 158 (Unit 1)Amendment No. 140 (Unit 2)

V1 9Page 3 of 3RCS Specific Activity3.4.16250IU-I.20015010050PERCENT OF RATED THERMAL POWERFIGURE 3.4.16-1REACTOR COOLANT DOSE EQUIVALENT I-131 REACTOR COOLANT SPECIFIC ACTIVITYLIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANTSPECIFIC ACITVITY

>1 mCi/gram DOSE EQUIVALENT 1-131Vogtle Units 1 and 23.4.16-4Amendment No. 96 (Unit 1)Amendment No. 74 (Unit 2)

V20Page 1 of 1RCS Operational LEAKAGE3.4.133.4 REACTOR COOLANT SYSTEM (RCS)3.4.13 RCS Operational LEAKAGELCO 3.4.13RCS operational LEAKAGE shall be limited to:a. No pressure boundary LEAKAGE;b. I1 gpm unidentified LEAKAGE;c I1 p dniidLAAE nd. 150 galosper idaytprimdLAryGE toscndar EKG hog none steam generator (SG).APPLICABILITY:

MODES 1, 2, 3, and 4.ACTIONS__________________

___CONDITION REQUIRED ACTION COMPLETION TIMEA. RCS operational A.1I Reduce LEAKAGE to 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />sLEAKAGE not within within limits.limits for reasons otherthan pressure boundaryLEAKAGE or primary tosecondary LEAKAGE.B. Required Action and B.1 Be in MODE 3. 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />sassociated Completion Time of Condition A not ANDmet.B.2 Be in MODE 5. 36 hour4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />sO__RPressure boundaryLEAKAGE exists.ORPrimary to secondary LEAKAGE not withinlimit.Vogtle Units 1 and 23.4.13-1Amendment No. 144 (Unit 1)Amendment No. 124 (Unit 2)

V21IPage 1 of 2Approved ByPoede VrinJ. B. Stanley Vogtle Electric Generating Plant 19200cdr 24.2i~Effective Date -0CTIASAEYFNTOSAUSRES Page Number7/25/12 F 0 C I CA SA E Y F N T O ST T S R ES9 of 11Sheet 1 of 1F- 0.5CONTAINMENT GOaTO19251-C.-'- PRESURELS u ==s4j TANji sl jI* a O T19251-(;IJ i .AT LEAST ONESCONTAINMENT SSPRAY PUMPSRUNNINGNOYES*egoGO TO* il ) 19261-CIGO TO19252-Cb/T> GO TO: ......1 9 2 6 3 -CCSP SAT,-r'nneu rebruary iZUll at 14:zz V21Page 2 of 2S0uthern Nuclear Operating CompanyA~rlN Plant: VEGP ! "X6CNA1 5ISUHda Unit: 1&2 Title: NEI 99-01 Rev 6 EAL Calculations SHEET 43If Containment pressure(PCTMT) exceeds the statichead (AH) dlue to thedifference between theTransfer Tube centerline elevation (EL 186"-93/4";

Design In puts #4 & #5) and PT~the SFP low operating water level (EL 21 Design Input #4), theTransfer Tube air-to-air barrier is not maintained."

AIH (ft) = 217'-0"-

186'-9.75"

= 30"-2.25"

= -30 ft(psig) > AH (ft) x p (Ibn/ft3) x g1 (ft/sec2) x 1 ft2go (Ibm -ft)/(lIbf-sec

2) 144 in2Pctmt (psig) > 30 ft x 61.55 Ibm x 32.2 (ft/sec2) x 1 ft2t332.2 (Ibm-ft)/(Ibf-sec

2) 144 in2(Design Input #25)Petmi > '43 psigPressure

> 52 psig WITH Tech Spec containment integrity intactNMP-EP-110-GL03 (page 8) defines CONTAINMENT INTEGRITY as The Primary Containment is OPERABLE per Technical Specification

3. 6.1.1." Tech Spec surveillance requirement

3. 6.1.1states "Perform required visual examinations and leakage rate testing except for containment air lock testing, in accordance with the Containment Leakage Rate Testing Program."

TechSpec section 5. 5.17 describes the Containment Leakage Rate Testing Program.

Per Tech SpecBases B3. 6.1, the Containment is designed to contain radioactive material that may be releasedfrom the reactor core following a Design Basis Accident (DBA).Per section 3.1.3.1 of VEGP Design Criteria DC-2 101, the mechanical (piping) and electrical penetrations, in conjunction with the carbon steel liner, form a leak-tight barrier.

Thus, thesepenetrations must meet the design accident pressure requirement of section 3. 4.5 of DC-2 101,52 psig.The absence of air-to-air containment penetrations during Modes 5 and 6 in VEGP procedure 142 10-1 was confirmed by e-mails from John Stanley (VEGP Operations Outage Manager; seeAttachment CS) and Ron Cowen (Westinghouse Site Services Manager; see Attachment C6).