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
Text
V9 Page 1 of 5 Southern Nuclear Operating Company I rlll Plant: VEGP TteNE990Re6EACacliosI X6CNA1 5 Unit: 1&2 TteNE9-0Re6ELCacliosSHEET 10 Volume 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 5.3.3-1 Hot Leg centerline elevation 187'-0" AX4DR023, 1X4DL4A017-1, &(76% RVLIS) 2X4DL4A01 7-1 Cold Leg centerline elevation Hot Leg Nozzle Bottom 185'-91/2' AX4DR023 Top 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 15.0.3-3 RCS operating pressure 2250 psia Full power coolant mass 2.53E+08 g Page 3 of LTR-CRA-06-1 79 attached to WEC-SNC letter GP-1 8006 and Table 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 0000-0 12. 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 V9 Page 2 of 5 Southern Nuclear Operating Company SOI AmiE Plant: VEGP ITitle: NEI 99-01 Rev 6 EAL Calculations X6CNA15~I Unit: 1&2 SHEET 37 CA1: Loss of RPV inventory.
Operating Mode Applicability:
Emergency Action Levels: Cold Shutdown, Refueling 1 OR2 1. Loss of RPV inventory as indicated be level less than elevation 185'-10" (73% on Full Range RVLIS).The RPV water level elevations corresponding to the RCS loop piping bottom IDs are found as follows: Dimension IElevation Loop Centerline Elevation 1 87'-00" Cold Leg Inside Diameter 27.5"1/2AxID 13.75" Bottom ID = Centerline
-(1/2Ax ID) 185'-1 0.25" Hot Leg Inside 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 water level 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 linearly interpolating between the TOAF (EL 181'-10" or 63% RVLIS) and the CL and HL centerline elevation (EL 187"0O" or 76% RVLIS):
V9 Page 3 of 5 Southern Nuclear Operating Cornpany ouIu A i= Plant: VEGP ITteNE990Re6EACacliosX6CNA1 5 I I ULOMPA Unit: 1&2 Til:NI9-Rv6ELCluaion SHEET 38 VEGP RVLIS Indication vs. RPV Water Level Elevation} .i 181 182 F 18 8 8 8 88 RPV 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 longer AND b. UNPLANNED increase in Containment sump, Reactor Coolant Drain Tank (RCDT)or Waste Holdup Tank (WHT) levels due to a loss of RPV inventory.
V9 Page 4 of 5 Southern Nuclear Operating Company 4xnlmM. Plant: VEGP ITitle: NEI 99-01 Rev 6 EAL Calculations X 6CNA15 I SO mUTH t Unit: 1&2 S HEET 39 CSI: Loss of RPV inventory affecting core decay heat removal capability.
Operating Mode Applicability:
Emergency Action Levels: Cold Shutdown, Refueling 1 OR20OR3 1. a. CONTAINMENT CLOSURE not established AND b. RPV water level less than 1 85'-4" [6" below Bottom ID of loop] (72% on Full Range RVLIS).The RPV water level elevations corresponding to 6" below the cold leg (CL) and hot leg (HL) bottom IDs are found as follows: Dimension Elevation Loop Centerline Elevation 1 87'-00" Cold Leg Inside 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 Leg Inside 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 linearly interpolating between the TOAF (EL 181 '-10" or 63% RVLIS) and the CL and HL centerline elevation (EL 187"-0" or 76% RVLIS):
V9 Page 5 of 5 Southern Nuclear Operating CompanyPlnt: VEGP Title: NEI 99-01 Rev 6 EAL CalculationsX6N1 mpw Unit: 1&2 I SHEET 40 VEGP RVLIS Indication vs. RPV Water Level Elevation 621 "B~owRCS-
__ ___.... .... ..... i .jPiping Bottom U)181 182 183 184 185 188 187 188 RPV Water Level Elevation (feet)The RPV water level elevation corresponding to 6" below the Bottom ID is 185'-4" or -72% on Full Range RVLIS.2. a. CONTAINMENT CLOSURE established AND b. RPV level less than 181'-1 0" ITOAF] (63% on Full Range RVLIS).3. a. RPV level cannot be monitored for 30 minutes or longer AND b. Core uncovery is indicated by ANY of the following:
RE-005 O..R 006 > 40 REM/hr Erratic Source Range monitor indication UNPLANNED increase in Containment Sump, Reactor Coolant Drain Tank (RCDT) or Waste Holdup Tanks (WHT) levels of sufficient magnitude to indicate core uncovery Vl 0 VEGP-FSAR-1 1Pae1o5 11.2.1.3 Eqluipment Design The LWPS equipment design parameters are provided in table 11.2.1-2.The seismic design classification and safety classification for the LWPS components and structures are listed in table 3.2.2-1. Safety class designations are also indicated on the LWPS piping 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 Water Reactors," 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 analyze fluids before discharge.
Based on the laboratory analysis, these wastes are either retained for further 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 analyses of known volumes of effluent.The radioactive liquid discharged from the reactor coolant system (RCS) is processed by the radwaste processing facility systems and may be discharged or recycled.The LWPS is arranged to recycle reactor grade water if desired. This is implemented by the segqregqation 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 from sources inside the containment.
B. Drain Channel A This portion of the LWPS collects aerated, reactor grade effluent that can be recycled.C. Drain Channel B This 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 portable demineralizers installed in subterranean enclosures inside the radwaste processing facility.
The radwaste processing facility is described in paragraph 11.4.2.4.
The radwaste processing facility demineralizers can be aligned to process any of the three waste drain streams.E. The radwaste processing facility filtration system consists of a portable, vendor supplied 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 radwaste 11.2-4 11.2-4REV 13 4106 Vl0a Page 2 of 5 VEGP-FSAR-1 1 processing facility.
The peripheral equipment is located adjacent to the filter assembly.
The filter system can be aligned to process any of the three waste drain 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 exchange resins.The LWPS does not include provisions for processing secondary system wastes. Secondary system effluent is handled by the steam generator blowdown processing system (SGBPS), as described in subsection 10.4.8, and by the turbine building drain system. Estimated releases from 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 by discharging low activity wastes directly, where permissible, with no treatment.
Instrumentation and controls necessary for the operation of the LWPS are located on a control board 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, and AX4DB1 24-5 and process flow diagram for the LWPS is shown on figure 11.2.2-1.
Table 11.2.1-1 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 as shown 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 section 11.1, 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 coolant Ipump No. 2 seal leakoffs, reactor vessel flange leakoff, and other deaerated, tritiated water Isources inside the containment.
Connections are provided for draining the RCS loops and the safety injection system (SIS) accumulators and for cooling the pressurizer relief tank. In addition, refueling canal drains can be routed to the refueling water storage tank using the RCDT pumps.The RCDT contents are continuously recirculated through the RCDT heat exchanger to maintain the desired temperature.
Level is prevented from varying significantly by a control valve which automatically opens a path from the recirculation line to the BRS when normal tank level 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 hydrogen blanket. Maintaining a constant level minimizes the amount of gas sent to the GWPS and minimizes the amount of hydrogen used. Provisions for sampling the gas are provided.Details of the RCDT subsystem are shown on drawing 1X4DB127.
A separate RCDT subsystem 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 holdup tank. Sources of this aerated liquid are as follows: A. Accumulator drainage (via RCDT pump suction).11.2-5 11.2-5REV 13 4/06 V10o Page 3 of 5 VEGP-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 or the 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 facility demineralizers before reuse or discharge.
Details of this equipment are shown on drawings AX4DB1 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 water with some radioactivity.
A separate drain channel A subsystem is provided for each of the two units. Details are shown on drawings 1X4DB124 and 1X4DB127.
Table 11.2.1-1 lists the estimated flows entering the waste holdup tank.11.2.2.3 Drain Channel B Subsystem Drain channel B is provided to collect and process nonreactor grade liquid wastes. These include:* 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 the following tanks.A. Laundry and Hot Shower Tank The laundry and hot shower tank is provided to collect and process waste effluents from the plant laundry and personnel decontamination showers and hand sinks.Laundry and hot shower drains normally need no treatment for removal of radioactivity.
This water is transferred to a waste monitor tank through the laundry and hot shower tank filter for eventual discharge.
If sample analysis indicates that decontamination is necessary, the water can be directed through the 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. Details of this portion of the LWPS are shown on drawing 1X4DB126.
Table 11.2.1-1 lists estimated flows entering the laundry and hot shower tank.B. Floor Drain Tank Water may enter the floor drain tank from system leaks inside the containment through the containment sump, from system leaks in the auxiliary building through auxiliary building sumps and the floor drains, and floor drains in the 11.2-6 11.2-6REV 13 4/06 v10o Page 4 of 5 VEGP-FSAR-1 1 radwaste 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 waste holdup tank.Another source of water to the floor drain tank is the chemical laboratory drains.Excess nonreactor grade samples that are not chemically contaminated and laboratory equipment rinse water are drained to the floor drain tank.The contents of the floor drain tank are processed through the radwaste processing 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 cannot be met without cleanup, the liquid can be processed by the waste monitor tank demineralizer, the radwaste processing facility demineralizers, or the radwaste processing facility filtration system.A separate floor drain tank and associated equipment are provided for each of the two units. Details of this portion of the LWPS are shown on drawing 1X4DB126.
Table 11.2.1-1 lists the estimated flows entering the floor drain tank.C. Chemical Drain Tank Laboratory samples which contain reagent chemicals (and possibly tritiated liquid) are discarded through a sample room sink which drains to the chemical drain tank. Chemical drains requiring radwaste processing are sent to the solid waste management system or may be processed through the radwaste processing 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.
Table 11.2.1-1 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.
The findings are logged, and, if the activity level is within acceptable limits, the tank contents are released 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 a distance sufficient to close the valve before passing the fluid that activated the detector trip signal. The isolation valve also blocks flow if sufficient dilution water is not available.
The radiation monitor is described in section 11.5. A permanent record of the radioactive releases is provided by a sample analysis of the known volumes of waste effluent released.
Liquid waste 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 to allow activity to decay to acceptable levels, or it can be further processed by the waste monitor 11.2-7 11.2-7REV 13 4/06 V10 Page 5 of 5 VEGP-FSAR-11I H. Waste Monitor Tank Pumps Two pumps are provided for each unit. One pump is used for each monitor tank to discharge water from the LWPS or for recycling if further processing is required.The pump may also be used for circulating the water in the waste monitor tank to obtain uniform tank contents, and therefore a representative sample, before discharge.
These pumps can be throttled to achieve the desired discharge rate.I. Auxiliary Waste Monitor Tank Pumps Two pumps are provided.
They are installed in Unit 2 but serve both units. One pump is used for each auxiliary waste monitor tank to discharge water from LWPS or for recycling if further processing is required.
A mixer may be used for circulating the water in the auxiliary waste monitor tank to obtain uniform tank contents, thereby assuring a representative sample is acquired prior to discharge of the tank contents.
The pumps can be throttled to achieve the desired discharge rate.11.2.2.6.2 Tanks A. Reactor Coolant Drain Tank One tank is provided for each unit. The purpose of the RCDT is to collect leakoff-type drains inside the containment at a central collection point for further disposition through a single penetration via the RCDT pumps. The tank provides surge volume and net positive suction head (NPSH) to the pumps.Only water which can be directed to the boron recycle holdup tanks enters the RCDT. The water is compatible with reactor coolant and does not contain dissolved air during normal plant operation, by engineering design.A constant level is maintained in the tank to minimize the amount of gas sent to the 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 valve in the discharge line. This valve operates on a signal from a level controller to limit the flow out of the system. The remainder of the flow is recirculated to the tank.Continuous flow is maintained through the heat exchanger in order to prevent loss of pump NPSH resulting from a sudden inflow of hot liquid into the RCDT.B. Waste Holdup Tank One 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 during normal operation.
C. Waste Evaporator Condensate Tank 11.2-11 11.2-11REV 13 4/06 Vii Page 1 of 3 Southern Nuclear Operating Company smrllllM~LPlant:
VEGP Title: NEI 99-01 Rev 6 EAL Calculations I 6CNA15¢m Unit: 1&2 SHEET 42 UNPLANNED increase in Containment Sump, Reactor Coolant Drain Tank (RCDT) or Waste Holdup Tanks (WHT) level of sufficient magnitude to indicate core uncovery.AND c. 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 in the RPV with an RPV water level of less than TOAF (18 1'-10" or 63% on Full Range RVLIS). 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, among them the requirement that >23' of water (EL 21 7'-0") is maintained above the RPV flange. This corresponds to the SEP water low level (Design Input #4).During Refueling Operations, fuel is moved between the Containment and the Fuel Handling Building via the Fuel Transfer Tube. The Fuel Transfer Tube will maintain an air-to-air barrier during 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.
Vll Page 2 of 3 Southern Nuclear Operating Cornpany AOm~I 4 Plant: VEGP Title: NEI 99-01 Rev 6 EAL Calculations I X6CNA1 5 I MV Unit: 1&2 I SHEET 53 The results of the Loss of Clad FP Barrier setpoint calculations in Attachments H-3 and 13 are summarized below. Given the system accuracy -a factor of two over the operating range -the threshold 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+07 VEGP 2 1.49E+04 1.5E+07 Containment Barrier Potential Loss Threshold 4.B Containment 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 the containment barrier.Sheet 23 of VEGP SAMG calculation X6CNA II established the 6% by volume hydrogen limit.
Vll Page 3 of 3 Desis,,C.lulation
-Nuclear Southern Conmpay Services Aim~Prjc:Vogtl lei Gener~atig Plat Ci.No. X6CNAn 1 5 SSubJectflitle:
Severe Accident Management Guideline (SAMG) Calculations Sheet 23 of 167 CA-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 //s WOO Severe Accident Management Guidance., Rev. 1 2. EpRI TR-101 869, Severe Acodient Managemnent Guidance Technical Basis Report, Volume 2: The Physics of Accident Progression
- 3. FSAR: a. Secion 6.2.1.5.2
- c. Figure 6.2.1-1 IN b. Table 6.2.5-6 d. Figur 6.2.1-4 4. Technical Specifications:
- a. Section 3.6.1.4 b. Section 3.6.1.5 1/5. Memo from Roger Hayes (PRA) on MAAP Case: MAAP 02-002-V (CO/CO 2 Results from Vogtle RPV Rupture Case), November 5, 2002 (copy attached on page 24)6. ASME Steam Tables, Fifth Edition Assumptions:
- 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 (wet percentage), 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 EXCEL spreadsheets.
A. The value of CO and CO 2 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 is shown on page 24.B. The methods used to determine the hydrogen flammability limits in containment are based on the WOO Severe Accident Guidelines (Ref. 1). The equations used are taken directly from these documents and are repeated below, along with any required design input values.
V12 Page 1 of 17 Southern Nuclear Operating Cornpany SOI AliM Plant: VEGP ITteNt990Re6EACaclios X6CNA1 5 I CMPANY Unit: 1&2 Til:NI9-1RvELCluaion SHEET 42 I UNPLANNED increase in Containment Sump, Reactor Coolant Drain Tank (RCDT) or Waste Holdup Tanks (WHT) level of sufficient magnitude to indicate core uncovery.AND c. 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 in the RPV with an RPV water level of less than TOAF (18 1'-10" or 63% on Full Range RVLIS). 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, among them the requirement that >23' of water (EL 21 7"0") is maintained above the RPV flange. This corresponds to the SEP water low level (Design Input #4).During Refueling Operations, fuel is moved between the Containment and the Fuel Handling Building via the Fuel Transfer Tube. The Fuel Transfer Tube will maintain an air-to-air barrier during 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.
Vi12 Page 2 of 17 Southern Nuclear Operating CompanyUnit: 1&2 Title: NEI 99-01 Rev 6 EAL Calculations SHEET 43 If Containment pressure (PCTMT) exceeds the static head (AtH) due to the difference between the Transfer Tube centerline elevation (EL 186"-93/4";Design Inputs #4 & #5) and P'm the SEP low operating water level (EL 217'-O"; H Design Input #4), the Transfer Tube air-to-air
..barrier is not maintained.
', AtH (ft) = 217'-0"- 186'-9.75" = 30"-2.25" = -30 ft Pctmt (psig) > AtH (ft) x p (Ibm/ft 3) x gt (ft/sec 2) x I ft 2 gc (Ibm-ft)/(Ibf-sec
- 2) 144 in 2 Pctmt (psig) > 30Oft x 61.551Ibm x 32.2 (ft/sec 2) x 1 ft 2 ft 3 32. 2 (Ibm-ft)/(Ibf-sec
- 2) 144 in 2 (Design Input #25)> -13 psig Pressure > 52 psig WITH Tech Spec containment integrity intact NMP-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 3.6.1.1 states "Perform required visual examinations and leakage rate testing except for containment air lock testing, in accordance with the Containment Leakage Rate Testing Program." Tech Spec section 5. 5.17 describes the Containment Leakage Rate Testing Program. Per Tech Spec Bases B3. 6.1, the Containment is designed to contain radioactive material that may be released from 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, these penetrations 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," see Attachment C5) and Ron Cowen (Westinghouse Site Services Manager; see Attachment C6).
V12 Page 3 of 17 Piping Penetrations The piping penetrations are listed in drawings 1X4DL 4A0 13, 1X4DL 4A014, 2X4DL 4A0 13, and 2X4DL 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 design temperature 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 not fail 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.
There are no operability or functionality issues because the peak containment DBA pressure is -37 psig (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 is 0. 01 cc/sec at DBA conditions.
CONTAINMENT CLOSURE no.t established.
Basis: NEI 99-01 Rev 6, page 81.
V12, Page 4 of17 Southern Nuclear Design Calculation SPlant: Vogtle Unit: 1&2 ICalculation Number: X6CNA15 ISheet: F-I Attachment F -Evaluation of 52 psig Pressure on Mechanical Penetrations Introduction There is a discrepancy between the DBA Design Pressures for the Containment (52 psig per section 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 pipe penetrations.
Conclusions The compressive and shear loads imposed by a 52 psig Containment pressure on the Unit 1&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 52 psig.Method A Type I pipe penetration is shown below: iQlrt & 4 N[From 1X4DL4A014
& 2X4DL4A014]
The weakest point of the penetration sleeve is the weld between the penetration sleeve and the containment liner. If the loads imposed by containment pressure on these welds are less than 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 thinner plate of the joint. The American Welding Society specifies allowable stresses of 20,000 psi in tension or compression and 13,600 psi in shear." The specifications for Containment liner welds are likely to be more stringent (i.e., higher allowable stresses) than the values in this textbook.
Using these textbook values is conservative for the purposes of this evaluation:
establishing an allowable limit.
V12, Page 5 of 17 Southern Nuclear Design Calculation IPlant: vogtle unit: 1&2 1Calculat°n Nubr: X6CNAI5 sheet: F-2 I Attachment F -Evaluation of 52 psig Pressure on Mechanical Penetrations These allowable stresses are most likely specified at standard temperature (68 F or 20 C). The maximum fluid temperature passing through one of these penetrations is 557 F (-290 F).Per VEGP FSAR Table 6.2.1-1, the peak DBA containment temperature is 250 F (-120 C). The yield strength of steel decreases with increasing temperature as shown in the representative graph to the right.Reducing the above allowable stresses by 15% conservatively addresses the effect of increased temperature 1,1 1.0 0,9 0,8 I-eI-U)0,75 0 200 400 600 Temperature
°C Variation 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 follows Lw=ix ID where ID = Inside diameter of penetration sleeve = OD -2 x t D = 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 t where 0 c-nom = Nominal allowable compressive stress (20,000 psi)ftemp = Reduction due to increased temperature
= 0.85 = 1 -0.15 Lw= 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 t where 0 s-nora = Nominal allowable shear stress (13,600 psi)fternp = Reduction due to increased temperature
= 0.85 = 1 -0.15 Lw= Weld length (inches)T = Weld thickness (inches) = Wall thickness (inches)
V12, Page 6 of 17 Southern 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 the penetration sleeve. Using the sleeve outside diameter (D in the above figure) maximizes this load: Pc = Pctmt x H x D 2/4 The Containment pressure (Pctmt psig) exerts a shear load (Ps Ibf) along the length of the penetration sleeve. Using the overall sleeve length (L in the above figure) maximizes this load: Ps = Pctmt x H- x D x L Evaluation The effect of a 52 psig Containment pressure on the Unit 1 and Unit 2 pipe penetrations are calculated in Excel spreadsheets Attachment F1 and Attachment F2.References F1. IX4DL4A0I3, Revision 7, "Containment Building Unit I Containment Wall Pipe Penetration Design List" F2. IX4DL4A014, Revision 9, "Containment Building Unit I Containment Wall Pipe Penetration Design List" F3. 2X4DL4A013, Revision 5, "Containment Building Unit 2 Containment Wall Pipe Penetration Design List" F4. 2X4DL4A014, Revision 4, "Containment Building Unit 2 Containment Wall Pipe Penetration Design List" F5. Singer, "Strength of Materials," second edition, 1962 V12, Page 7 of 17 X6CNAI5 ATTACHMENT F SHEET F-4 Bornt, Butch From: Jani, Yogendra M.Sent: Tuesday, October 14, 2014 4:52 PM To: Borer, Butch Cc: Patel, V. R.; Evans, William P. (SNC Corporate);
Lambert, David Leslie
Subject:
FW: VEGP Pipe Penetration Eval Butch, 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 of ASME 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 Jani SNC Fleet Des -Safety Anl & Mech 205.992.5125 office 205.410.9806 mobile V12, Page 8 of 17 SHEET Fl-i X6CNAI15 ATTACHMENT F1 Evaluation 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 ID ac = Allowable compressive stress (psi)a3c = a3c-nom X ftemp= 20,000 psi = Nominal allowable comprssive stress ftemp = 0.85 = Reduction due to increased temperature ac = 17,000 psi a's =a's =Allowable shear stress (psi)O's-fara X ftemp= 13,600 psi = Nominal allowable comprssive stress ftemp = 0.85 = Reduction due to increased temperature 11,560 psi Fc = O'c x Lw x t = Allowable Compressive Load (Ibf)Fs = as x Lw x t = Allowable Shear Load (Ibf)Pctmnt =52 psig = Containment Pressure P 0 = 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 17 SHEET F1-2 X6CNAI5 ATTACHMENT F1 Evaluation of 52 psig Pressure on U1 Pipe Penetrations Compression LI I= Tp V orVII peerto;dimension B used instead of L Type V Penetration:
Per IX4DL4A0 14, there is no dimension L; use B instead Type VII Penetration:
If dimension L not provided on 1X4DL4A013 or IX4DL4A014, use B instead PEN # Type L D t ID Lw Fc Pc PclFc 1-4 I 48.25 56.000 1.500 53.000 167 4.2E+06 1.3E+05 0.030 5 VII 8.250 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.024 7-10 I 18.000 18.000 0.500 17.000 53 4.5E+05 1.3E+04 0.029 11& 12 III 11.750 12.750 0.375 12.000 38 2.4E+05 6.6E+03 0.028 13 III 12.750 12.750 0.375 12.000 38 2.4E+05 6.6E+03 0.028 14 VII 15.250 10.750 0.365 10.020 31 2.OE+05 4.7E+03 0.024 15 II 15.750 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.024 16 -17 V 8.250 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.024 18-21 I 37.500 34.000 1.500 31.000 97 2.5E+06 4.7E+04 0.019 22 II 15.750 10.750 .0.365 10.020 31 2.0E+05 4.7E+03 0.024 23 I 14.930 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.024 24 II 15.750 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.024 25 V 8.250 18.000 0.500 17.000 53 4.5E+05 1.3E+04 0.029 28 &29 II 16.750 18.000 0.500 17.000 53 4.5E+05 1.3E+04 0.029 30 & 31 II 18.750 14.000 0.438 13.124 41 3.1E+05 8.0E+03 0.026 32 II 15.750 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.024 33 II 18.750 14.000 0.438 13.124 41 3.1E+05 8.0E+03 0.026 34 & 35 II 19.250 20.000 0.500 19.000 60 5.1E+05 1.6E+04 0.032 40 II 16.750 18.000 0.500 17.000 53 4.5E+05 1.3E+04 0.029 41 II 15.750 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.024 42 II 12.000 10.750 0.365 10.020 31 2.OE+05 4.7E+03 0.024 43-46 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 0.024 48 I 15.750 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.024 49 II 15.750 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.024 50 I 15.750 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.024 51 -55 I 15.750 10.750 0.365 10.020 31 2.OE+05 4.7E+03 0.024 V12, Page 10 of 17 SHEET F1-3 X6CNA15 ATTACHMENT F1 Evaluation of 52 psig Pressure on U1 Pipe Penetrations Compression
[111I= TyeV orViipntain dimension B used instead of L Type V Penetration:
Per 1X4DL4AO014, there is no dimension L; use B instead Type VII Penetration:
if dimension L not provided on 1X4DL4A013 or 1X4DL4AO14, use B instead PEN # Type JLJD~ t IDJ__ Fc_ Pc JPcIFc 56 I__44.750134.000I1.500 31.0001 97 2.5E+06 4.7E+04 0.019 57 & 58 I 32.000 24.000 1.000 22.000 69 1 .2E+06 2.4E+04 0.020 59 & 60 I 28.000 26.000 1.000 24.000 75 1 .3E+06 2.8E+I04 0.022 61 V 8.250 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.024 62 &63 II 15.750 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.024 64 VI 13.250 10.750 0.365 10.020 31 2.0E+05 4.7E+i03 0.024 66 V j8.250 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.024 67 III 12.750 12.750 0.375 12.000 38 2.4E+05 6.6E+03 0.028 68 II 16.000 18.000 0.500 17.000 53 4.5E+05 1.3E+04 0.029 69 -73 III 12.750 12.750 0.375 12.000 38 2.4E+05 6.6E+03 0.028 75 V 8.250 10.750 0.365 10.020 131 2.0E+'05 4.7E+03 0.024 76 V 8.250 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.024 77 & 78 II 15.750 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.024 79 II 12.000 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.024 80 II 16.000 18.000 0.500 17.000 53 4.5E+05 1.3E+'04 0.029 81 II 12.000 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.024 82 V 8.250 18.000 0.500 17.000 53 4.5E+05 1.3E+04 0.029 83 & 84 10.000 24.000 0.500 23.000 72 6.1E+05 2.4E+04 0.038 85 III 12.750 12.750 0.375 12.000 38 2.4E+05 6.6E+03 0.028 86 III 23.670 12.750 0.375 12.000 38 2.4E+05 6.6E+03 0.028 87 II 16.000 18.000 0.500 17.000 53 4.5E+05 1.3E+04 0.029 88 VII 15.250 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.024 90 VII 8.250 10.750 0.365 10.020 31 2.0E+05 4.7E+'03 0.024 91 -98 II 19.750 18.000 0.750 16.500 52 6.6E+05 1.3E+04 0.020 100 9.250 4.500 0.237 4.026 13 5.1E+04 8.3E+02 0.016 101 -104 I 20.550 18.000 0.750 16.500 52 6.6E+05 1.3E+04 0.020 Maximum PclFc = 0.038 V12, Page 11 of 17 SHEET FI-4 X6CNAI15 ATTACHMENT F1 Evaluation of 52 psig Pressure on UI Pipe Penetrations Shear[111= Type V or VII penetration; dimension B used instead of L Type V Penetration:
Per IX4DL4A0 14, there is no dimension L; use B instead Type VII Penetration:
If dimension L not provided on 1X4DL4A013 or 1X4DL4A014, use B instead PEN # IType L 0 t ID Lw Fs Ps JPs/Fs 1 -4 I 48.250 56.000 1.5001 53.0001167 2.9E+06 4.4E+05 0.153 5 VII 8.250 10.750 0.365 10.020 31.5 1.3E+05 1.4E+04 0.109 7- 10 I 18.000 18.000 0.500 17.000 53.4 3.1E+05 5.3E+04 0.171 11& 12 III 11.750 12.750 0.375 12.000 37.7 1.6E+05 2.4E+04 0.150 13 III 12.750 12.750 0.375 12.000 37.7 1.6E+05 2.7E+04 0.163 14 VII 15.250 10.750 0.365 10.020 31.5 1.3E+05 2.7E+04 0.202 15 II 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.208 16- 17 V 8.250 10.750 0.365 10.020 31.5 1.3E+05 1.4E+04 0.109 18-21 I 37.500 34.000 1.500 31.000 97.4 1.7E+06 2.1E+05 0.123 22 II 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.208 23 I 14.930 10.750 0.365 10.020 31.5 1.3E+05 2.6E+04 0.197 24 II 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.208 25 V 8.250 18.000 0.500 17.000 53.4 3.1E+05 2.4E+04 0.079 28 & 29 II 16.750 18.000 0.500 17.000 53.4 3.1E+05 4.9E+04 0.160 30 & 31 II 18.750 14.000 0.438 13.124 41.2 2.1E+05 4.3E+04 0.205 32 II 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.208 33 II 18.750 14.000 0.438 13.124 41.2 2.1E+05 4.3E+04 0.205 34 & 35 II 19.250 20.000 0.500 19.000 59.7 3.5E+05 6.3E+04 0.182 40 II 16.750 18.000 0.500 17.000 53.4 3.1E+05 4.9E+04 0.160 41 II 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.208 42 II 12.000 10.750 0.365 10.020 31.5 1.3E+05 2.1E+04 0.159 43-46 II 18.750 18.000 0.500 17.000 53.4 3.1E+05 5.5E+04 0.179 47 V 8.250 10.750 0.365 10.020 31.5 1.3E+05 1.4E+04 0.109 48 I 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.208 49 II 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.208 50 I 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.208 51 -55 I 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.208 V12, Page 12 of 17 SHEET Fl-5 X6CNA1 5 ATTACHMENT Fl Evaluation of 52 psig Pressure on UI Pipe Penetrations Shear I ... I=Type V or VII penetration; dimension B used instead of L Type V Penetration:
Per IX4DL4A0 14, there is no dimension L; use B instead Type VII Penetration:
If dimension L not provided on IX4DL4A013 or 1X4DL4A014, use B instead PEN # Type L D t ID Lw Fs Ps P 5/F 5 56 I 44.750 34.000 1.500 31.000 97.4 1.7E+06 2.5E+05 0.147 57 & 58 I 32.000 24.000 1.000 22.000 69.1 8.0E+05 1.3E+05 0.157 59 & 60 I 28.000 26.000 1.000 24.000 75.4 8.7E+05 1.2E+05 0.136 61 V o8.25'0 10.750 0.365 10.020 31.5 1.3E+05 1.4E+04 0.109 62 &63 II 15.750 10.750 0.365 10.020 31.5 i.3E+05 2.8E+04 0.208 64 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 0.109 67 III 12.750 12.750 0.375 12.000 37.7 1.6E+05 2.7E+04 0.163 68 II 16.000 18.000 0.500 17.000 53.4 3.1E+05 4.7E+04 0.152 69 -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 0.109 77 & 78 II 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.208 79 II 12.000 10.750 0.365 10.020 31.5 1.3E+05 2.IE+04 0.159 80 II 16.000 18.000 0.500 17.000 53.4 3.IE+05 4.7E+04 0.152 81 II 12.000 10.750 0.365 10.020 31.5 1.3E+05 2.1E+04 0.159 182 ....8.250 18.000 0.500 17.000 53.4 3.1E+05 2.4E+04 0.079 83 & 84 10.000 24.000 0.500 23.000 72.3 4.2E+05 3.9E+04 0.094 85 III 12.750 12.750 0.375 12.000 37.7 1.6E+05 2.7E+04 0.163 86 III 23.670 12.750 0.375 12.000 37.7 1.6E+05 4.9E+-04 0.302 87 II 16.000 18.000 0.500 17.000 53.4 3.IE+05 4.7E+04 0.152 88: 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 0.109 91 -98 II 19.750 18.000 0.750 16.500 51.8 4.5E+05 5.8E+04 0.129 100 9.250 4.500 0.237 4.026 12.6 3.5E+04 6.8E+03 0.196 101,-104 I 20.550 18.000 0.750 16.500 51.8 4.5E+05 6.0E+04 0.134 Maximum PslFs = 0.302 V12, Page 13 of 17 X6CNA15 ATTACHMENT F2 SHEET F2-1 Evaluation 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 ftemp ac-norn = 20,000 psi = Nominal allowable comprssive stress ftermp = 0.85 = Reduction due to increased temPerature O-c = 17,000 psi = Allowable compressive stress as = Allowable shear stress (psi)a's = a's-nom X ftemp a'c-nomn = ,I13,600 psi = Nominal allowable comprssive stress ftemp = 0.85 = RedUction due to increased temperature O's = 11,560 psi = Allowable shear stress Fc = ac x Lw x t = Allowable Compressive Load (Ibf)Fs = as x Lw x t = Allowable Shear Load (Ibf)Pctmnt = 52 psig = Containment Pressure Pc = 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 17 SHEET F2-2 X6CNA15S ATTACHMENT F2 Evaluation of 52 psig Pressure on U2 Pipe Penetrations Compression I I= Type V or VII penetration; dimension B used instead of L iType V Penetration:
Per 2X4DL4A0 14, there is no dimension L; use B instead Type VII Penetration:
If dimension L not provided on 2X4DL4A013 or 2X4DL4A014, use B instead PEN # Type L D t ID Lw FcPc PclFc 1 -4 I 48.25 56.000 1.500 53.000 167 4,2E+06 1.3E+05 0.030 5 VII 8.250 10.750 0.365 10.020 31 2,0E+05 4,7E+03 0.024 7- 10 I 18.000 18.000 0.500 17.000 53 4.5E+05 1,3E+04 0.029 11& 12 III 11.750 12.750 0.375 12.000 38 2,4E+05 6,6E+03 0.028 13 III 12.750 12.750 0.375 12.000 38 2.4E+05 6.6E+03 0.028 14 VII 15.250 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.024 15 II 15.750 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.024 16 -17 V 8.250 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.024 18-21 I 37.500 34.000 1.500 31.000 97 2.5E+06 4,7E+04 0.019 22 II 15.750 10.750 0.365 10.020 31 2,0E+05 4,7E+03 0.024 23 I 14.930 10.750 0.365 10.020 31 2.0E+05 4,7E+03 0.024 24 II 15.750 10.750 0.365 10.020 31 2,0E+05 4,7E+03 0.024 25 ,V 8.250 18.000 0.500 17.000 53 4.5E+05 1,3E+04 0.029 28 & 29 II 16.750 18.000 0.500 17.000 53 4.5E+05 1,3E+04 0.029 30 & 31 II 18.750 14.000 0.438 13.124 41 3.1E+05 8,0E+03 0.026 32 II 15.750 10.750 0.365 10.020 31 2,0E+05 4,7E+03 0.024 33 II 18.750 14.000 0.438 13.124 41 3,1E+05 8,0E+03 0.026 34 & 35 II 19.250 20.000 0.500 19.000 60 5.IE+05 1,6E+04 0.032 40 II 16.750 18.000 0.500 17.000 53 4,5E+05 1,3E+04 0.029 41 II 15.750 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.024 42 II 12.000 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.024 43 -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 0.024 48 I 15.750 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.024 49 II 15.750 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.024 50 I 15.750 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.024 51 -55 I 15.750 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.024 V12, Page 15 of 17 SHEET F2-3 X6CNA15S ATTACHMENT F2 Evaluation of 52 psig Pressure on U2 Pipe Penetrations Compression
~=TpeorI=
pentraIon dimension Bused instead of L Type V Penetration:
Per 2X4DL4A0 14, there is no dimension L; use B instead Type VII Penetration:
If dimension L not provided on 2X4DL4A013 or 2X4DL4A014, use B instead PEN # Type L D t ID LFcPc Pc/Fc 56 I 44.750 34.000 1.500 31.000 97 2.5E+06 4.7E+04 0.019 57 & 58 I 32.000 24.000 1.000 22.000 69 1.2E+06 2.4E+04 0.020 59 & 60 1 28.000 26.000 1.000 24.000 75 1 .3E+06 2.8E+04 0.022 61 V 8.250 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.024 62 &63 II 15.750 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.024 64 VI 13.250 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.024 66 V 8.250 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0,024 67 III 12,750 12,750 0.375 12,000 38 2,4E+05 6.6E+03 0,028 68 II 16.000 18.000 0.500 17.000 53 4.5E+05 1.3E+04 0.029 69-73 III 12.750 12.750 0.375 12.000 38 2.4E+05 6.6E+03 0.028 75 V 8.250 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.024 76 V 8.250 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.024 77 &78 II 15.750 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.024 79 II 12,000 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.024 80 II 16.000 18.000 0.500 17.000 53 4.5E+05 1.3E+04 0.029 81 II 12.000 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.024 82 V 8.250 18.000 0.500 17.000 53 4.5E+05 1.3E+04 0.029 83 &84 10.000 24.000 0.500 23.000 72 6.IE+05 2.4E+04 0.038 85 III 12.750 12.750 0.375 12.000 38 2.4E+05 6.6E+03 0.028 86 III 23.670 12.750 0.375 12.000 38 2.4E+05 6.6E+03 0.028 87 II 16.000 18.000 0.500 17.000 53 4.5E+05 1.3E+04 0.029 88 VII 15.250 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.024 90 VII 8.250 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.024 91 -98 II 19.750 18.000 0.750 16.500 52 6.6E+05 1.3E+04 0.020 100 9.250 4.500 0.237 4.026 13 5.1E+04 8.3E+02 0.016 101 -104 I 20.550 18.000 0.750 16.500 52 6.6E+05 1.3E+04 0.020 Maximum PdIFc = 0.038 V12, Page 16 of 17 SHEET F2-4 X6CNA15S ATTACHMENT F2 Evaluation of 52 psig Pressure on U2 Pipe Penetrations Shear Type V orVillpntain dimension B used instead of L Type V Penetration:
Per 2X4DL4A014, there is no dimension L; use B instead Type VII Penetration:
If dimension L not provided on 2X4DL4A013 or 2X4DL4A014, use B instead PEN # Type L D t ID Lw Fs Ps PslFs 1 -4 I 48.250 56.000 1.500 53.000 167 2.9E+06 4.4E+05 0.153 5 VII 8.250 10.750 0.365 10.020 31.5 1,3E+05 1.4E+04 0.109 7-10 I 18.000 18.000 0.500 17.000 53.4 3.1E+05 5.3E+04 0.171 11& 12 III 11.750 12.750 0.375 12.000 37.7 1.6E+05 2.4E+04 0.150 13 III 12.750 12.750 0.375 12.000 37.7 1.6E+05 2.7E+04 0.163 14 VII 15.250 10.750 0.365 10.020 31.5 1.3E+05 2.7E+04 0.202 15 II 15.750 10.750 !0.365 10.020 31.5 1.3E+05 2.8E+04 0.208 16-17 V 8.250 10.750 0.365 10.020 31.5 1.3E+05 1.4E+04 0.109 18-21 1 37.500 34.000 1.500 31.000 97.4 1.7E+06 2.IE+05 0.123 22 II 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.208 23 I 14.930 10.750 0.365 10.020 31.5 1.3E+05 2.6E+04 0.197 24 II 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.208 25 V 8.250 18.000 0.500 17.000 53.4 3.1E+05 2.4E+04 0.079 28 &29 II 16.750 18.000 0.500 17.000 53.4 3.1E+05 4.9E+04 0.160 30 & 31 II 18.750 14.000 0.438 13.124 41.2 2.1E+05 4.3E+04 0.205 32 II 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.208 33 II 18.750 14.000 0.438 13.124 41.2 2.IE+05 4.3E+04 0.205 34 & 35 II 19.250 20.000 0.500 19.000 59.7 3.5E+05 6.3E+04 0.182 40 II 16.750 18.000 0.500 17.000 53.4 3.IE+05 4.9E+04 0.160 41 II. 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.208 42 II 12.000 10.750 0.365 10.020 31.5 1.3E+05 2.1E+04 0.159 43-46 II 18.750 18.000 0.500 17.000 53.4 3.1E+05 5.5E+04 0.179 47 V 8.250 10.750 0.365 10.020 31.5 1.3E+05 1.4E+04 0.109 48 I 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.208 49 II 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.208 50 I 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.208 51-55 I 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.208 V12, Page 17 of 17 SHEET F2-5 X6CNA1 5 ATTACHMENT F2 Evaluation of 52 psi 9 Pressure on U2 Pipe Penetrations Shear[111= Type V or VII penetration; dimension B used instead of L Type V Penetration:
Per 2X4DL4A014, there is no dimension L; use B instead Type VII Penetration:
If dimension L not provided on 2X4DL4A013 or 2X4DL4A014, use B instead PEN # Type L D t ID L FsPs PslFs 56 I 44.750 34.000 1.500 31.000 97.4 1.7E+06 2.5E+05 0.147 57 & 58 I 32.000 24.000 1.000 22.000 69.1 8.0E+05 1.3E+05 0,157 59 & 60 I 28.000 26.000 1.000 24.000 75.4 8.7E+05 1.2E+05 0.136 61 V 8.250 10.750 0.365 10.020 31.5 1.3E+05 1.4E+04 0.109 62 &63 II 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.208 64 VI 13.250 10.750 0.365 10.020 31.5 1.3E+05 2.3E+04 0.175 66 V 8.250 10.750 0.365 10.020 31.5 1.3E+05 1.4E+04 0.109 67 III 12.750 12.750 0.375 12.000 37.7 1.6E+05 2.7E+04 0.163 68 II 16.000 18.000 0.500 17.000 53.4 3.1E+05 4.7E+04 0.152 69-73 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 76 V 8.250, 10.750 0.365 10.020 31.5 1.3E+05 1.4E+04 0.109 77 &78 II 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.208 79 II 12.000 10.750 0.365 10.020 31.5 1.3E+05 2.1E+04 0.159 80 II 16.000 18.000 0.500 17.000 53.4 3.1E+05 4.7E+04 0.152 81 II 12.000 10.750 0.365 10.020 31.5 1.3E+05 2.1E+04 0.159 82 V 8.250 18.000 0.500 17.000 53.4 3.1E+05 2.4E+04 0.079 83 & 84 10.000 24.000 0.500 23.000 72.3 4.2E+05 3.9E+04 0.094 85 III 12.750 12.750 0.375 12.000 37.7 1.6E+05 2.7E+04 0,163 86 III 23.670 12.750 0.375 12.000 37.7 1.6E=+05 4.9E=+04 0.302 87 II 16.000 18.000 0.500 17.000 53.4 3.1E+05 4.7E+04 0.152 88 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 0.109 91 -98 II 19.750 18.000 0.750 16.500 51.8 4.5E+05 5.8E+04 0.129 100 9.250 4.500 0.237 4.026 12.6 3.5E+04 6.8E+03 0.196 101 -104 I 20.550 18.000 0.750 16.500 51.8 4.5E+05 6.0E+-04 0.134 Maximum PslFs 0.302 I-+/- 4.4.4 .J~XH~_ ~ Th.7 r~.j Vi3 Page 1 of 1 T<-H........ .....--I ALVI W. VOCTHE P1AJ hC.;a i i a i V1 4 Page 1 of 2 provided 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 located inside battery rooms.3.2 POWER GENERATION DESIGN BASES None 3.3 MAJOR COMPONENT DESIGN BASES 3.3.1 The Class IE dc 125-V dc system and all equipment are classified as Seismic Category 1 and 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 voltage range of 105 to 14OV dc.The 25 kVA inverters provide power to RIHR isolation valves. The breakers for these valves are located in "Trip" position during normal plant operation; they do not operate during the duty cycle time of the battery, or for any LOCA/LOOP or SBO coping scenario.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 required loads for a loss-of-coolant accident, loss of offsite power, or main steam line break for a duration 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 battery replacement 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 6VR1 6 VER 13 V14 Page 2 of 2 3.3.3.7 A dc ammeter, dc voltmeter, dc overvoltage relay, ac power "on" light, and ac undervoltage relay shall be provided on each charger. The relays shall alarm in the control 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 to recharge a discharged battery in a shorter amount of time (if allowed by the manufacturer).
3.3.3.9 The battery charger shall prevent the charger from becoming a load on the battery due to a 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 the charger from the ac source.3.3.3.11 dc breakers shall be provided to protect the battery and charger from internal charger faults 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 vital instrumentation 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 shall also 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-phase inverter 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 105 to 140 V dc. The dc feeder cables shall be designed to maintain a minimum of 105 V dc during 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 to 140 V dc. The dc feeder cables to the RHR isolation valve inverters shall be designed to maintain a minimum of 105 V dc at the RH-R isolation valve inverters.
3.3.5 125-V dc MCCs 3.3.5.1 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 8VR1 8 VER 13 X6CNA1 5 Attachment L
C2I.l~at~inn fnr I:W.I I1 SHEET L-6 v15 Page 1 of 3 CALC NO. SNC024-CALC-004 VEGP DETERMINATION OF EMERGENCY ACTION LEVEL REV. 0 I0 E N E R C 0 N FOR INITIATING CONDITION E-PAGE NO. Page 6of 8 5.0 Design Inputs 1. The contact dose rates from the HI-STORM 100 and HI-TRAC 125 cask system technical specification
[2, Table 6.2-3] are provided below in Table 5-1. These source 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-STORM 100 and HI-TRAC 125 Number of Technical Specification LoatonjMeasurements j Limit (mrem/hr)HI-TRAC 1.25 __________
Side -Mid -height 4 472.7 Top 1 4 j102.4 HI-STORM 100 ____ _____Side -60 inches below mid-height 4 87 Side -Mid -height 4 88.9 Side -60 inches above mid-height 4 54.8 Top -Center of lid 1 24.5 Top -Radially centered 4 29.2 Inlet duct 4 178.8 Outlet duct 4 64.5 X6CNA1 5 Attachment L ENERCON Calculation for E-HU1 SHEET L-7 v15 Page 2 of 3 CALC NO. SNC024-CALC-004 VEGP DETERMINATION OF ____________
EMERGENCY ACTION LEVEL REV. 0 0 E N E R C 0 N FOR INITIATING CONDITION E- -__________
PAGE NO. Page 7of 8 6.0 Methodology The "on-contact" dose rates from the technical specification for the HI STORM-i100 cask system are scaled by a factor of 2, as specified in NEI 99-01 Rev. 6 [1], for use in initiating condition E-HUI.
X6CNA1 5 Attachment L FNFRf.ON ('nlrmiitinn fnr F-HI- 11 SHEET L-8V1 Page 3 of 3 CALC NO. SNC024-CALC-004 VEGP DETERMINATION OF -____________
EMERGENCY ACTION LEVEL REV. 0 0 E N E R C 0 N FOR INITIATING CONDITION E-PAGE NO. Page 8of 8 7.0 Calculations The dose rates in Table 5-1 are multiplied by 2 in order to calculate the EAL dose rate limits. These calculations are presented below in Table 7-1.Table 7-1 Dose Rate Scaling Calculations for EAL Limits Technical LoainSpecification Scaling Calculated Value EAL LoainLimit Factor (mrem/hr) (mrem/hr)_________________________ (mrem/hr) j____ _______________
HI-TRAC 125___ _____Side -Mid -height [ 472.7 2 1 945.4 [ 950 Top 102.4 2 j 204.8 [ 200 HI-STORM 100 Side -60 inches below mid-height 87 2 174 170 Side -Mid -height 88.9 2 177.8 180 Side -60 inches above mid-height 54.8 2 109.6 110 Top -Center of lid 24.5 2 49 50 Top -Radially centered 29.2 2 58.4 60 Inlet duct 178.8 2 357.6 360 Outlet duct 64.5 2 129 130 8.0 Computer Software Microsoft WORD 2013 is used in this calculation for basic multiplication.
V1 6 Page 1 of 4 Approved By J. B. Stanley Effective Date 7/25/12~Voqgtle Electric Generating F-0 CRITICAL SAFETY FUNCTION Si F- 0.2 CORE COOLING PatProcedure version Pat19200-C 24.2 Page Number'ATUS TREES5ol Sheet 1 of 1-4 GO TO 11221-V~GO TO 19221-C VLI5 FULL NO GE GREATER tHAN 41% ,E8S* .GO TO 19222-C 11.FI GO TO 19222Z-C tVUI FULL N iGE GREATER THAN 41% YE.GO TO* 19223-V~GO TO rYNAMIC Ha 9,Z-INGE N ho -4 RCP h-3 RCP h-2RCP E-1IRCF* 19223-C GI CSA T v Printed February 2, 2016 at 16:09 Vi16 Page 2 of 4 Approved By Procedure Veso J. B. Stanley Vogtle Electric Generating P..lantng 19200-C Veso24.2 Effective Date Page Number 7/25/12 F-0 CRITICAL SAFETY FUNCTION STATUS TREES6of1 Sheet 1 of 1 F- 0.3 HEAT SINK I KTOTAL AVAILABLE FEEDWATER FLOW TO SGs GREATER iTHAN 570 GPM NO YES NARROW RANGE LEVEL IN AT LEAST rONE SG GREATER THAN 10% (32%)GO TO 1 9231 -C GO TO 19232-C GO TO 1 9233-C NO PRESSURE IN ALL SGs LESS THAN 1240 PSIG YE LESS THAN 82% YES NO PRESSURE IN ALL SGs LESS THAN 1180 PSIG YES GO TO 1 9234-C GO TO 19235-C NARROW RANGE NO LEVEL IN ALL SGs GREATER THAN 10% (32%) YES L SAT Printed February 2, 2016 at 16:09 V1 6 Page 3 of 4 Approved By Procedure Version J. B. Stanley Vogtle Electric Generating Plant 19200-C 24.2 Effective Date Page Number 7/25/12 F-0 CRITICAL SAFETY FUNCTION STATUS TREES7of1 Sheet 1 of 1 F- 0.4 INTEGRITY GO TO 19241-C---- .) GO TO I 19241-C ALL RCS WR COLD LEG TEMPERATURES GREATER THAN 295 0 F~QGO TO 19242-C~*CSF SAT TEMPERATURE DECREASE IN ALL NO RCS COLD LEGS LESS100 0 F IN THE ILAST iYES!60 MINUTES JGOTO I,11 19241-C ALL RCS WR COLD LEG NO TEMPERATURES GREATER THAN 265°F YES RCS PRESSURE LESS NO THAN COLD OVERPRESSURE LIMIT 465 PSIG YES ,I RCS WR COLD LEG NO TEMPERATURE GREATER THAN 220 0 F YES LL* (I'\ GO TO" "" 19242-C O CSF SAT-*CSF SAT Printed February 2, 2016 at 16:09 Vi16 Page 4 of 4 Approved By .....Procedure Version J. B. StanI~yVogtle Electric Generating Plant '19200-C 2.Effe tiv Datan ey P... ..age... :.. .... .... ... ...... Numbe Efeciv DteF-0 CRITICAL SAFETY FUNCTION STATUS TREES 9ag ofmb1r 7/25/129of1 Sheet 1 of 1 F- 0.5 CONTAINMENT GO TO 19251-C* )GO TO i 19251-C N.E N o P GO TO p 19252-C I CONTAINMENT BUMP LEVEL LESS THAN 196 iNCHESGO TO.......19253.C Ib CSF SAT Printed February 2, 2016 at 16:09 V1 7 Page 1 of 3 Approved By p, 4Pocdr Version W. L. Burmeister
... Vogtle Electric Generating Plant IS55039-c~rcdr 3.2 Effective Date P~age Number 0510712013 ISEISMIC MONITORING INSTRUMENTATION SYSTEM I 6 of 9 4.2 NORMAL OPERATION NONE 4.3 NON-PERIODIC OPERATIONS NOTE This subsection shall be initiated by 50022-C "Seismic Event Plan" or Supervisor's direction.
1 4.3.1 Retrieving Seismic Data NOTE SKey Number 1-OP3-10 is in C&T. LI a. Verify the Event alarm on the Condor Control Unit screen is RED. El b. Obtain the event charts and graphs from each recorder.
El 4.3.2 Retrieving Seismic Data At ETNA (River Intake)a. Hook up a laptop to the uplink cable. El b. Power on computer, THEN click on the ALTUS Quick Talk icon. El c. In the ALTUS Status window, verify the alarm has been triggered.
El d. 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 recorded event El (3) Click on the Retrieve File button El (4) Download the file to a floppy disk El Printed February 15, 2016 at 16:59 V17 Page 2 of 3 Seismic Event Plan 50022-C VOGTLE Version 14.0 Unit C Page 7of 24 4.0 INSTRUCTION 4.1 IDENTIFICATION OF SEISMIC EVENT 1. The following indicators are available for determining whether or not seismic event 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 has occurred; (Must be called when any of the above indications of an earthquake is received) and to initiate this procedure if earthquake is confirmed.
- 2. IF an earthquake is sensibly detected by control room personnel, initiate this procedure.
- 3. IF AOP 18036-C, Seismic Event, is in effect, then initiate this procedure.
Printed February 15, 2016 at 17:02 V1 7 Page 3 of 3 Seismic Event 18036-C VOGTLE Version 11 Unit C Page 4of 9 PURPOSE The purpose of this procedure is to provide operator response following a seismic event and to initiate 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 VI18 Page 1 of 5 Southern Nuclear Design Calculation IPlant: Vogte Unit: 1&2 Icalculation Number: X6CNAI4 Isheet:;46 Miscellaneous Design Inputs 21. 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, which are 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 r Isotope Average Gamma Energy (MeV)
Reference:
Brookhaven National Laboratory National Nuclear Data Center decay data (http://www.orau .or qlptp/PTP%20Libraryllibrary/DOE/bnl/nu clidedata/table.htm)
Copies of web pages in Attachment C8 1-131 0.382 1-132 2.20 1-133 0.607 I-134 2.50 1-135 1.55 Co-580.975 Co-60 2.51 Cs-134 1.55 Cs-136 2.12 Cs-i137 0.565 Cs-138 2.31 Cs-138 2.31 Vi18 Page 2 of 5 Southern Nuclear Design Calculation SPlant: Vogtle Unit: 1&2 ICalculation Number: X6CNA14 ISheet: 61 Recognition Category S: System Malfunctions Notice of Unusual Event SU4: Fuel Clad Degradation.
Operating Mode Applicability:
Power Operation (Mode 1)Startup (Mode 2)Hot Standby (Mode 3)Hot Shutdown (Mode 4)1 OR2 Emergency Action Levels: SU4 EALI: CVCS Letdown radiation monitor RE-48000 reading greater than 5 pCi/cc indicating 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/g Per section B.3.4.16, page B3.4.16-2 of VEGP Tech Spec Bases, noble gas activity in the reactor coolant assumes 1% failed fuel, which closely equals the 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 13 of X6AZ01 A, the principle isotopes detected by RE-48000 are 1-131, 1-133, Co-58, Co-60, Cs-134, and Cs-137.However, per Section B-12-3-2 and Figure B-12-2 of 1X6AZ01-10004 & 2X6AZ01-10004, RE-48000 will detect gammas of energies down to-0.1 MeV.S t 1 __ _ "__I-, -.. I _ _ _ _i c -i ..L =t 4 ..II; P.m IENKR4Y It ,VgL= t.VI Figure B-12-2 Thus the other I, Co, and Cs isotopes listed in FSAR Table 11.1-2 should be included if their average decay gamma energies exceed 0.1 MeV.
V1 8 Page 3 of 5 Southern Nuclear Design Calculation iPlant: Vogtle U nit: 1&2 ICalculation Number: X6CNAI4 Sheet: 62 Per LTR-CRA-06-179 attached to WEC-SNC letter GP-18006, the pre-MURPU coolant activities may be adjusted upward 2% to account for the increase in core thermal power from 3565 MWt to 3636 MWt. Thus, the Co and Cs MURPU 1% defect activity are equal to their pre-MURPU 1% Defect activities multiplied by 1.02.The Co and Cs activities corresponding to the 1.0 DE 1-131 Tech Spec limit are the products of their MURPU 1% defect activities and the ratio 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 the CVCS 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% Defect Isotope Coolant Coolant Coolant Activity Activity Activity I-131 0.74 2.91 ______I-132 0.75 2.96 ______I-133 1.41 5.56 1-134 0.18 0.69 ______I-135 0.69 2.72 ______Co-58 3.89E-03 1 .53E-02 1 .50E-02 Co-60 4.93E-04 1 .94E-03 1 .90E-03 Cs-134 5.97E-01 2.35 2.3 Cs-I136 7.52E-01 2.96 2.9 Cs-137 3.89E-01 1.53 1.5 Total = 5.5 21.7 ptCi/g Total = 5.5 21.5 i.LCi/cc CVCS Letdown Density =0.99 g/cc SGiven 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 8 Southern Nuclear Design Calculation Page 4 of 5 SPlant: Vogtle Unit: 1&2 Calculation Number: X6CNA14 Sheet: C5-1 Attachment C5 -VEGP 1&2 CVCS Letdown Radiation Monitor (RE-48000)
Readings U 1.... ... .a IIII-~' I~~-:-~a..
V1 8 Southern Nuclear Design Calculation Page 5 of 5 SPlant: Vogtle Unit: 1&2 Calculation Number: X6CNA14 Sheet: C5-2 Attachment C5 -VEGP 1&2 CVCS Letdown Radiation Monitor (RE-48000)
Readings*]W11 o=I~IMY Iin'~vu l~u~r ~.~tImP~.ii I~' ~'~I WE ~'~~jL 11U WOWW4m~ ~
V19 Page 1 of 3 RCS Specific Activity 3.4.16 3.4 REACTOR COOLANT SYSTEM (RCS)3.4.16 RCS Specific Activity LCO 3.4.16 APPLICABILITY:
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--------------------------
INlJ LCO 3.0.4c is applicable.
I------------------
---CONDITION REQUIRED ACTION COMPLETION TIME A. DOSE EQUIVALENT A.1 Verify DOSE Once per4 hours I-131 > 1.0 p.Ci/gm. EQUIVALENT I-131 within the acceptable region of Figure 3.4.16-1.AND A.2 Restore DOSE 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> EQUIVALENT I-131 to within limit.B. Gross specific activity of B.1 Perform SR 3.4.16.2.
4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> the reactor coolant not within limit. AND 8.2 Be in MODE 3 with 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Tavg < 500°F.(continued)
Vogtle Units 1 and 2 3.4.16-1 Amendment No. 137 (Unit 1)Amendment No. 116 (Unit 2)
V1 9 Page 2 of 3 RCS Specific Activity 3.4.16 ACTIONS (continued)
________________
__________
CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and C.1 Be in MODE 3 with 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Tavg < 500°F.Time of Condition A not met.O_.R DOSE EQUIVALENT 1-131 in the unacceptable region of Figure 3.4.16-1.SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.16.1 Verify reactor coolant gross specific In accordance with activity_
100/I. !iCi/gm. the Surveillance Frequency Control Program SR 3.4.16.2 ---- --NOTE- --- --Only required to be performed in MODE 1.Verify reactor coolant DOSE EQUIVALENT I-131 In accordance with specific activity < 1.0 ,.tCi/gm, the Surveillance Frequency Control Program AND Between 2 and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after a THERMAL POWER change of _> 15% RTP within 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 2 3.4.16-2 Amendment No. 158 (Unit 1)Amendment No. 140 (Unit 2)
V1 9 Page 3 of 3 RCS Specific Activity 3.4.16 250 I U-I.200 150 100 50 PERCENT OF RATED THERMAL POWER FIGURE 3.4.16-1 REACTOR COOLANT DOSE EQUIVALENT I-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACITVITY >1 mCi/gram DOSE EQUIVALENT 1-131 Vogtle Units 1 and 2 3.4.16-4 Amendment No. 96 (Unit 1)Amendment No. 74 (Unit 2)
V20 Page 1 of 1 RCS Operational LEAKAGE 3.4.13 3.4 REACTOR COOLANT SYSTEM (RCS)3.4.13 RCS Operational LEAKAGE LCO 3.4.13 RCS operational LEAKAGE shall be limited to: a. No pressure boundary LEAKAGE;b. I1 gpm unidentified LEAKAGE;c I1 p dniidLAAE n d. 150 galosper idaytprimdLAryGE toscndar EKG hog n one steam generator (SG).APPLICABILITY:
MODES 1, 2, 3, and 4.ACTIONS__________________
___CONDITION REQUIRED ACTION COMPLETION TIME A. RCS operational A.1I Reduce LEAKAGE to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> LEAKAGE not within within limits.limits for reasons other than pressure boundary LEAKAGE or primary to secondary LEAKAGE.B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition A not AND met.B.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> O__R Pressure boundary LEAKAGE exists.OR Primary to secondary LEAKAGE not within limit.Vogtle Units 1 and 2 3.4.13-1 Amendment No. 144 (Unit 1)Amendment No. 124 (Unit 2)
V21I Page 1 of 2 Approved ByPoede Vrin J. B. Stanley Vogtle Electric Generating Plant 19200cdr 24.2i~Effective Date -0CTIASAEYFNTOSAUSRES Page Number 7/25/12 F 0 C I CA SA E Y F N T O ST T S R ES9 of 11 Sheet 1 of 1 F- 0.5 CONTAINMENT GOaTO 19251-C.-'- PRESURELS u ==s4 j TANji sl j I* a O T 19251-(;I J i .AT LEAST ONE SCONTAINMENT SSPRAY PUMP SRUNNING NO YES*ego GO TO* il ) 19261-C I GO TO 19252-C b/T> GO TO: ......1 9 2 6 3 -C CSP SAT ,-r'nneu rebruary iZUll at 14:zz V21 Page 2 of 2 S0uthern Nuclear Operating Company A~rlN Plant: VEGP ! "X6CNA1 5 ISUHda Unit: 1&2 Title: NEI 99-01 Rev 6 EAL Calculations SHEET 43 If Containment pressure (PCTMT) exceeds the static head (AH) dlue to the difference between the Transfer 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), the Transfer 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/ft 3) x g1 (ft/sec2) x 1 ft 2 go (Ibm -ft)/(lIbf-sec
- 2) 144 in 2 Pctmt (psig) > 30 ft x 61.55 Ibm x 32.2 (ft/sec2) x 1 ft 2 t 3 32.2 (Ibm-ft)/(Ibf-sec
- 2) 144 in 2 (Design Input #25)Petmi > '43 psig Pressure > 52 psig WITH Tech Spec containment integrity intact NMP-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. 6.1.1 states "Perform required visual examinations and leakage rate testing except for containment air lock testing, in accordance with the Containment Leakage Rate Testing Program." Tech Spec section 5. 5.17 describes the Containment Leakage Rate Testing Program. Per Tech Spec Bases B3. 6.1, the Containment is designed to contain radioactive material that may be released from 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, these penetrations 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; see Attachment CS) and Ron Cowen (Westinghouse Site Services Manager; see Attachment C6).
V9 Page 1 of 5 Southern Nuclear Operating Company I rlll Plant: VEGP TteNE990Re6EACacliosI X6CNA1 5 Unit: 1&2 TteNE9-0Re6ELCacliosSHEET 10 Volume 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 5.3.3-1 Hot Leg centerline elevation 187'-0" AX4DR023, 1X4DL4A017-1, &(76% RVLIS) 2X4DL4A01 7-1 Cold Leg centerline elevation Hot Leg Nozzle Bottom 185'-91/2' AX4DR023 Top 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 15.0.3-3 RCS operating pressure 2250 psia Full power coolant mass 2.53E+08 g Page 3 of LTR-CRA-06-1 79 attached to WEC-SNC letter GP-1 8006 and Table 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 0000-0 12. 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 V9 Page 2 of 5 Southern Nuclear Operating Company SOI AmiE Plant: VEGP ITitle: NEI 99-01 Rev 6 EAL Calculations X6CNA15~I Unit: 1&2 SHEET 37 CA1: Loss of RPV inventory.
Operating Mode Applicability:
Emergency Action Levels: Cold Shutdown, Refueling 1 OR2 1. Loss of RPV inventory as indicated be level less than elevation 185'-10" (73% on Full Range RVLIS).The RPV water level elevations corresponding to the RCS loop piping bottom IDs are found as follows: Dimension IElevation Loop Centerline Elevation 1 87'-00" Cold Leg Inside Diameter 27.5"1/2AxID 13.75" Bottom ID = Centerline
-(1/2Ax ID) 185'-1 0.25" Hot Leg Inside 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 water level 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 linearly interpolating between the TOAF (EL 181'-10" or 63% RVLIS) and the CL and HL centerline elevation (EL 187"0O" or 76% RVLIS):
V9 Page 3 of 5 Southern Nuclear Operating Cornpany ouIu A i= Plant: VEGP ITteNE990Re6EACacliosX6CNA1 5 I I ULOMPA Unit: 1&2 Til:NI9-Rv6ELCluaion SHEET 38 VEGP RVLIS Indication vs. RPV Water Level Elevation} .i 181 182 F 18 8 8 8 88 RPV 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 longer AND b. UNPLANNED increase in Containment sump, Reactor Coolant Drain Tank (RCDT)or Waste Holdup Tank (WHT) levels due to a loss of RPV inventory.
V9 Page 4 of 5 Southern Nuclear Operating Company 4xnlmM. Plant: VEGP ITitle: NEI 99-01 Rev 6 EAL Calculations X 6CNA15 I SO mUTH t Unit: 1&2 S HEET 39 CSI: Loss of RPV inventory affecting core decay heat removal capability.
Operating Mode Applicability:
Emergency Action Levels: Cold Shutdown, Refueling 1 OR20OR3 1. a. CONTAINMENT CLOSURE not established AND b. RPV water level less than 1 85'-4" [6" below Bottom ID of loop] (72% on Full Range RVLIS).The RPV water level elevations corresponding to 6" below the cold leg (CL) and hot leg (HL) bottom IDs are found as follows: Dimension Elevation Loop Centerline Elevation 1 87'-00" Cold Leg Inside 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 Leg Inside 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 linearly interpolating between the TOAF (EL 181 '-10" or 63% RVLIS) and the CL and HL centerline elevation (EL 187"-0" or 76% RVLIS):
V9 Page 5 of 5 Southern Nuclear Operating CompanyPlnt: VEGP Title: NEI 99-01 Rev 6 EAL CalculationsX6N1 mpw Unit: 1&2 I SHEET 40 VEGP RVLIS Indication vs. RPV Water Level Elevation 621 "B~owRCS-
__ ___.... .... ..... i .jPiping Bottom U)181 182 183 184 185 188 187 188 RPV Water Level Elevation (feet)The RPV water level elevation corresponding to 6" below the Bottom ID is 185'-4" or -72% on Full Range RVLIS.2. a. CONTAINMENT CLOSURE established AND b. RPV level less than 181'-1 0" ITOAF] (63% on Full Range RVLIS).3. a. RPV level cannot be monitored for 30 minutes or longer AND b. Core uncovery is indicated by ANY of the following:
RE-005 O..R 006 > 40 REM/hr Erratic Source Range monitor indication UNPLANNED increase in Containment Sump, Reactor Coolant Drain Tank (RCDT) or Waste Holdup Tanks (WHT) levels of sufficient magnitude to indicate core uncovery Vl 0 VEGP-FSAR-1 1Pae1o5 11.2.1.3 Eqluipment Design The LWPS equipment design parameters are provided in table 11.2.1-2.The seismic design classification and safety classification for the LWPS components and structures are listed in table 3.2.2-1. Safety class designations are also indicated on the LWPS piping 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 Water Reactors," 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 analyze fluids before discharge.
Based on the laboratory analysis, these wastes are either retained for further 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 analyses of known volumes of effluent.The radioactive liquid discharged from the reactor coolant system (RCS) is processed by the radwaste processing facility systems and may be discharged or recycled.The LWPS is arranged to recycle reactor grade water if desired. This is implemented by the segqregqation 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 from sources inside the containment.
B. Drain Channel A This portion of the LWPS collects aerated, reactor grade effluent that can be recycled.C. Drain Channel B This 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 portable demineralizers installed in subterranean enclosures inside the radwaste processing facility.
The radwaste processing facility is described in paragraph 11.4.2.4.
The radwaste processing facility demineralizers can be aligned to process any of the three waste drain streams.E. The radwaste processing facility filtration system consists of a portable, vendor supplied 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 radwaste 11.2-4 11.2-4REV 13 4106 Vl0a Page 2 of 5 VEGP-FSAR-1 1 processing facility.
The peripheral equipment is located adjacent to the filter assembly.
The filter system can be aligned to process any of the three waste drain 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 exchange resins.The LWPS does not include provisions for processing secondary system wastes. Secondary system effluent is handled by the steam generator blowdown processing system (SGBPS), as described in subsection 10.4.8, and by the turbine building drain system. Estimated releases from 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 by discharging low activity wastes directly, where permissible, with no treatment.
Instrumentation and controls necessary for the operation of the LWPS are located on a control board 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, and AX4DB1 24-5 and process flow diagram for the LWPS is shown on figure 11.2.2-1.
Table 11.2.1-1 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 as shown 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 section 11.1, 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 coolant Ipump No. 2 seal leakoffs, reactor vessel flange leakoff, and other deaerated, tritiated water Isources inside the containment.
Connections are provided for draining the RCS loops and the safety injection system (SIS) accumulators and for cooling the pressurizer relief tank. In addition, refueling canal drains can be routed to the refueling water storage tank using the RCDT pumps.The RCDT contents are continuously recirculated through the RCDT heat exchanger to maintain the desired temperature.
Level is prevented from varying significantly by a control valve which automatically opens a path from the recirculation line to the BRS when normal tank level 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 hydrogen blanket. Maintaining a constant level minimizes the amount of gas sent to the GWPS and minimizes the amount of hydrogen used. Provisions for sampling the gas are provided.Details of the RCDT subsystem are shown on drawing 1X4DB127.
A separate RCDT subsystem 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 holdup tank. Sources of this aerated liquid are as follows: A. Accumulator drainage (via RCDT pump suction).11.2-5 11.2-5REV 13 4/06 V10o Page 3 of 5 VEGP-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 or the 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 facility demineralizers before reuse or discharge.
Details of this equipment are shown on drawings AX4DB1 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 water with some radioactivity.
A separate drain channel A subsystem is provided for each of the two units. Details are shown on drawings 1X4DB124 and 1X4DB127.
Table 11.2.1-1 lists the estimated flows entering the waste holdup tank.11.2.2.3 Drain Channel B Subsystem Drain channel B is provided to collect and process nonreactor grade liquid wastes. These include:* 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 the following tanks.A. Laundry and Hot Shower Tank The laundry and hot shower tank is provided to collect and process waste effluents from the plant laundry and personnel decontamination showers and hand sinks.Laundry and hot shower drains normally need no treatment for removal of radioactivity.
This water is transferred to a waste monitor tank through the laundry and hot shower tank filter for eventual discharge.
If sample analysis indicates that decontamination is necessary, the water can be directed through the 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. Details of this portion of the LWPS are shown on drawing 1X4DB126.
Table 11.2.1-1 lists estimated flows entering the laundry and hot shower tank.B. Floor Drain Tank Water may enter the floor drain tank from system leaks inside the containment through the containment sump, from system leaks in the auxiliary building through auxiliary building sumps and the floor drains, and floor drains in the 11.2-6 11.2-6REV 13 4/06 v10o Page 4 of 5 VEGP-FSAR-1 1 radwaste 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 waste holdup tank.Another source of water to the floor drain tank is the chemical laboratory drains.Excess nonreactor grade samples that are not chemically contaminated and laboratory equipment rinse water are drained to the floor drain tank.The contents of the floor drain tank are processed through the radwaste processing 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 cannot be met without cleanup, the liquid can be processed by the waste monitor tank demineralizer, the radwaste processing facility demineralizers, or the radwaste processing facility filtration system.A separate floor drain tank and associated equipment are provided for each of the two units. Details of this portion of the LWPS are shown on drawing 1X4DB126.
Table 11.2.1-1 lists the estimated flows entering the floor drain tank.C. Chemical Drain Tank Laboratory samples which contain reagent chemicals (and possibly tritiated liquid) are discarded through a sample room sink which drains to the chemical drain tank. Chemical drains requiring radwaste processing are sent to the solid waste management system or may be processed through the radwaste processing 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.
Table 11.2.1-1 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.
The findings are logged, and, if the activity level is within acceptable limits, the tank contents are released 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 a distance sufficient to close the valve before passing the fluid that activated the detector trip signal. The isolation valve also blocks flow if sufficient dilution water is not available.
The radiation monitor is described in section 11.5. A permanent record of the radioactive releases is provided by a sample analysis of the known volumes of waste effluent released.
Liquid waste 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 to allow activity to decay to acceptable levels, or it can be further processed by the waste monitor 11.2-7 11.2-7REV 13 4/06 V10 Page 5 of 5 VEGP-FSAR-11I H. Waste Monitor Tank Pumps Two pumps are provided for each unit. One pump is used for each monitor tank to discharge water from the LWPS or for recycling if further processing is required.The pump may also be used for circulating the water in the waste monitor tank to obtain uniform tank contents, and therefore a representative sample, before discharge.
These pumps can be throttled to achieve the desired discharge rate.I. Auxiliary Waste Monitor Tank Pumps Two pumps are provided.
They are installed in Unit 2 but serve both units. One pump is used for each auxiliary waste monitor tank to discharge water from LWPS or for recycling if further processing is required.
A mixer may be used for circulating the water in the auxiliary waste monitor tank to obtain uniform tank contents, thereby assuring a representative sample is acquired prior to discharge of the tank contents.
The pumps can be throttled to achieve the desired discharge rate.11.2.2.6.2 Tanks A. Reactor Coolant Drain Tank One tank is provided for each unit. The purpose of the RCDT is to collect leakoff-type drains inside the containment at a central collection point for further disposition through a single penetration via the RCDT pumps. The tank provides surge volume and net positive suction head (NPSH) to the pumps.Only water which can be directed to the boron recycle holdup tanks enters the RCDT. The water is compatible with reactor coolant and does not contain dissolved air during normal plant operation, by engineering design.A constant level is maintained in the tank to minimize the amount of gas sent to the 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 valve in the discharge line. This valve operates on a signal from a level controller to limit the flow out of the system. The remainder of the flow is recirculated to the tank.Continuous flow is maintained through the heat exchanger in order to prevent loss of pump NPSH resulting from a sudden inflow of hot liquid into the RCDT.B. Waste Holdup Tank One 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 during normal operation.
C. Waste Evaporator Condensate Tank 11.2-11 11.2-11REV 13 4/06 Vii Page 1 of 3 Southern Nuclear Operating Company smrllllM~LPlant:
VEGP Title: NEI 99-01 Rev 6 EAL Calculations I 6CNA15¢m Unit: 1&2 SHEET 42 UNPLANNED increase in Containment Sump, Reactor Coolant Drain Tank (RCDT) or Waste Holdup Tanks (WHT) level of sufficient magnitude to indicate core uncovery.AND c. 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 in the RPV with an RPV water level of less than TOAF (18 1'-10" or 63% on Full Range RVLIS). 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, among them the requirement that >23' of water (EL 21 7'-0") is maintained above the RPV flange. This corresponds to the SEP water low level (Design Input #4).During Refueling Operations, fuel is moved between the Containment and the Fuel Handling Building via the Fuel Transfer Tube. The Fuel Transfer Tube will maintain an air-to-air barrier during 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.
Vll Page 2 of 3 Southern Nuclear Operating Cornpany AOm~I 4 Plant: VEGP Title: NEI 99-01 Rev 6 EAL Calculations I X6CNA1 5 I MV Unit: 1&2 I SHEET 53 The results of the Loss of Clad FP Barrier setpoint calculations in Attachments H-3 and 13 are summarized below. Given the system accuracy -a factor of two over the operating range -the threshold 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+07 VEGP 2 1.49E+04 1.5E+07 Containment Barrier Potential Loss Threshold 4.B Containment 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 the containment barrier.Sheet 23 of VEGP SAMG calculation X6CNA II established the 6% by volume hydrogen limit.
Vll Page 3 of 3 Desis,,C.lulation
-Nuclear Southern Conmpay Services Aim~Prjc:Vogtl lei Gener~atig Plat Ci.No. X6CNAn 1 5 SSubJectflitle:
Severe Accident Management Guideline (SAMG) Calculations Sheet 23 of 167 CA-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 //s WOO Severe Accident Management Guidance., Rev. 1 2. EpRI TR-101 869, Severe Acodient Managemnent Guidance Technical Basis Report, Volume 2: The Physics of Accident Progression
- 3. FSAR: a. Secion 6.2.1.5.2
- c. Figure 6.2.1-1 IN b. Table 6.2.5-6 d. Figur 6.2.1-4 4. Technical Specifications:
- a. Section 3.6.1.4 b. Section 3.6.1.5 1/5. Memo from Roger Hayes (PRA) on MAAP Case: MAAP 02-002-V (CO/CO 2 Results from Vogtle RPV Rupture Case), November 5, 2002 (copy attached on page 24)6. ASME Steam Tables, Fifth Edition Assumptions:
- 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 (wet percentage), 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 EXCEL spreadsheets.
A. The value of CO and CO 2 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 is shown on page 24.B. The methods used to determine the hydrogen flammability limits in containment are based on the WOO Severe Accident Guidelines (Ref. 1). The equations used are taken directly from these documents and are repeated below, along with any required design input values.
V12 Page 1 of 17 Southern Nuclear Operating Cornpany SOI AliM Plant: VEGP ITteNt990Re6EACaclios X6CNA1 5 I CMPANY Unit: 1&2 Til:NI9-1RvELCluaion SHEET 42 I UNPLANNED increase in Containment Sump, Reactor Coolant Drain Tank (RCDT) or Waste Holdup Tanks (WHT) level of sufficient magnitude to indicate core uncovery.AND c. 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 in the RPV with an RPV water level of less than TOAF (18 1'-10" or 63% on Full Range RVLIS). 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, among them the requirement that >23' of water (EL 21 7"0") is maintained above the RPV flange. This corresponds to the SEP water low level (Design Input #4).During Refueling Operations, fuel is moved between the Containment and the Fuel Handling Building via the Fuel Transfer Tube. The Fuel Transfer Tube will maintain an air-to-air barrier during 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.
Vi12 Page 2 of 17 Southern Nuclear Operating CompanyUnit: 1&2 Title: NEI 99-01 Rev 6 EAL Calculations SHEET 43 If Containment pressure (PCTMT) exceeds the static head (AtH) due to the difference between the Transfer Tube centerline elevation (EL 186"-93/4";Design Inputs #4 & #5) and P'm the SEP low operating water level (EL 217'-O"; H Design Input #4), the Transfer Tube air-to-air
..barrier is not maintained.
', AtH (ft) = 217'-0"- 186'-9.75" = 30"-2.25" = -30 ft Pctmt (psig) > AtH (ft) x p (Ibm/ft 3) x gt (ft/sec 2) x I ft 2 gc (Ibm-ft)/(Ibf-sec
- 2) 144 in 2 Pctmt (psig) > 30Oft x 61.551Ibm x 32.2 (ft/sec 2) x 1 ft 2 ft 3 32. 2 (Ibm-ft)/(Ibf-sec
- 2) 144 in 2 (Design Input #25)> -13 psig Pressure > 52 psig WITH Tech Spec containment integrity intact NMP-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 3.6.1.1 states "Perform required visual examinations and leakage rate testing except for containment air lock testing, in accordance with the Containment Leakage Rate Testing Program." Tech Spec section 5. 5.17 describes the Containment Leakage Rate Testing Program. Per Tech Spec Bases B3. 6.1, the Containment is designed to contain radioactive material that may be released from 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, these penetrations 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," see Attachment C5) and Ron Cowen (Westinghouse Site Services Manager; see Attachment C6).
V12 Page 3 of 17 Piping Penetrations The piping penetrations are listed in drawings 1X4DL 4A0 13, 1X4DL 4A014, 2X4DL 4A0 13, and 2X4DL 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 design temperature 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 not fail 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.
There are no operability or functionality issues because the peak containment DBA pressure is -37 psig (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 is 0. 01 cc/sec at DBA conditions.
CONTAINMENT CLOSURE no.t established.
Basis: NEI 99-01 Rev 6, page 81.
V12, Page 4 of17 Southern Nuclear Design Calculation SPlant: Vogtle Unit: 1&2 ICalculation Number: X6CNA15 ISheet: F-I Attachment F -Evaluation of 52 psig Pressure on Mechanical Penetrations Introduction There is a discrepancy between the DBA Design Pressures for the Containment (52 psig per section 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 pipe penetrations.
Conclusions The compressive and shear loads imposed by a 52 psig Containment pressure on the Unit 1&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 52 psig.Method A Type I pipe penetration is shown below: iQlrt & 4 N[From 1X4DL4A014
& 2X4DL4A014]
The weakest point of the penetration sleeve is the weld between the penetration sleeve and the containment liner. If the loads imposed by containment pressure on these welds are less than 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 thinner plate of the joint. The American Welding Society specifies allowable stresses of 20,000 psi in tension or compression and 13,600 psi in shear." The specifications for Containment liner welds are likely to be more stringent (i.e., higher allowable stresses) than the values in this textbook.
Using these textbook values is conservative for the purposes of this evaluation:
establishing an allowable limit.
V12, Page 5 of 17 Southern Nuclear Design Calculation IPlant: vogtle unit: 1&2 1Calculat°n Nubr: X6CNAI5 sheet: F-2 I Attachment F -Evaluation of 52 psig Pressure on Mechanical Penetrations These allowable stresses are most likely specified at standard temperature (68 F or 20 C). The maximum fluid temperature passing through one of these penetrations is 557 F (-290 F).Per VEGP FSAR Table 6.2.1-1, the peak DBA containment temperature is 250 F (-120 C). The yield strength of steel decreases with increasing temperature as shown in the representative graph to the right.Reducing the above allowable stresses by 15% conservatively addresses the effect of increased temperature 1,1 1.0 0,9 0,8 I-eI-U)0,75 0 200 400 600 Temperature
°C Variation 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 follows Lw=ix ID where ID = Inside diameter of penetration sleeve = OD -2 x t D = 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 t where 0 c-nom = Nominal allowable compressive stress (20,000 psi)ftemp = Reduction due to increased temperature
= 0.85 = 1 -0.15 Lw= 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 t where 0 s-nora = Nominal allowable shear stress (13,600 psi)fternp = Reduction due to increased temperature
= 0.85 = 1 -0.15 Lw= Weld length (inches)T = Weld thickness (inches) = Wall thickness (inches)
V12, Page 6 of 17 Southern 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 the penetration sleeve. Using the sleeve outside diameter (D in the above figure) maximizes this load: Pc = Pctmt x H x D 2/4 The Containment pressure (Pctmt psig) exerts a shear load (Ps Ibf) along the length of the penetration sleeve. Using the overall sleeve length (L in the above figure) maximizes this load: Ps = Pctmt x H- x D x L Evaluation The effect of a 52 psig Containment pressure on the Unit 1 and Unit 2 pipe penetrations are calculated in Excel spreadsheets Attachment F1 and Attachment F2.References F1. IX4DL4A0I3, Revision 7, "Containment Building Unit I Containment Wall Pipe Penetration Design List" F2. IX4DL4A014, Revision 9, "Containment Building Unit I Containment Wall Pipe Penetration Design List" F3. 2X4DL4A013, Revision 5, "Containment Building Unit 2 Containment Wall Pipe Penetration Design List" F4. 2X4DL4A014, Revision 4, "Containment Building Unit 2 Containment Wall Pipe Penetration Design List" F5. Singer, "Strength of Materials," second edition, 1962 V12, Page 7 of 17 X6CNAI5 ATTACHMENT F SHEET F-4 Bornt, Butch From: Jani, Yogendra M.Sent: Tuesday, October 14, 2014 4:52 PM To: Borer, Butch Cc: Patel, V. R.; Evans, William P. (SNC Corporate);
Lambert, David Leslie
Subject:
FW: VEGP Pipe Penetration Eval Butch, 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 of ASME 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 Jani SNC Fleet Des -Safety Anl & Mech 205.992.5125 office 205.410.9806 mobile V12, Page 8 of 17 SHEET Fl-i X6CNAI15 ATTACHMENT F1 Evaluation 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 ID ac = Allowable compressive stress (psi)a3c = a3c-nom X ftemp= 20,000 psi = Nominal allowable comprssive stress ftemp = 0.85 = Reduction due to increased temperature ac = 17,000 psi a's =a's =Allowable shear stress (psi)O's-fara X ftemp= 13,600 psi = Nominal allowable comprssive stress ftemp = 0.85 = Reduction due to increased temperature 11,560 psi Fc = O'c x Lw x t = Allowable Compressive Load (Ibf)Fs = as x Lw x t = Allowable Shear Load (Ibf)Pctmnt =52 psig = Containment Pressure P 0 = 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 17 SHEET F1-2 X6CNAI5 ATTACHMENT F1 Evaluation of 52 psig Pressure on U1 Pipe Penetrations Compression LI I= Tp V orVII peerto;dimension B used instead of L Type V Penetration:
Per IX4DL4A0 14, there is no dimension L; use B instead Type VII Penetration:
If dimension L not provided on 1X4DL4A013 or IX4DL4A014, use B instead PEN # Type L D t ID Lw Fc Pc PclFc 1-4 I 48.25 56.000 1.500 53.000 167 4.2E+06 1.3E+05 0.030 5 VII 8.250 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.024 7-10 I 18.000 18.000 0.500 17.000 53 4.5E+05 1.3E+04 0.029 11& 12 III 11.750 12.750 0.375 12.000 38 2.4E+05 6.6E+03 0.028 13 III 12.750 12.750 0.375 12.000 38 2.4E+05 6.6E+03 0.028 14 VII 15.250 10.750 0.365 10.020 31 2.OE+05 4.7E+03 0.024 15 II 15.750 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.024 16 -17 V 8.250 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.024 18-21 I 37.500 34.000 1.500 31.000 97 2.5E+06 4.7E+04 0.019 22 II 15.750 10.750 .0.365 10.020 31 2.0E+05 4.7E+03 0.024 23 I 14.930 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.024 24 II 15.750 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.024 25 V 8.250 18.000 0.500 17.000 53 4.5E+05 1.3E+04 0.029 28 &29 II 16.750 18.000 0.500 17.000 53 4.5E+05 1.3E+04 0.029 30 & 31 II 18.750 14.000 0.438 13.124 41 3.1E+05 8.0E+03 0.026 32 II 15.750 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.024 33 II 18.750 14.000 0.438 13.124 41 3.1E+05 8.0E+03 0.026 34 & 35 II 19.250 20.000 0.500 19.000 60 5.1E+05 1.6E+04 0.032 40 II 16.750 18.000 0.500 17.000 53 4.5E+05 1.3E+04 0.029 41 II 15.750 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.024 42 II 12.000 10.750 0.365 10.020 31 2.OE+05 4.7E+03 0.024 43-46 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 0.024 48 I 15.750 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.024 49 II 15.750 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.024 50 I 15.750 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.024 51 -55 I 15.750 10.750 0.365 10.020 31 2.OE+05 4.7E+03 0.024 V12, Page 10 of 17 SHEET F1-3 X6CNA15 ATTACHMENT F1 Evaluation of 52 psig Pressure on U1 Pipe Penetrations Compression
[111I= TyeV orViipntain dimension B used instead of L Type V Penetration:
Per 1X4DL4AO014, there is no dimension L; use B instead Type VII Penetration:
if dimension L not provided on 1X4DL4A013 or 1X4DL4AO14, use B instead PEN # Type JLJD~ t IDJ__ Fc_ Pc JPcIFc 56 I__44.750134.000I1.500 31.0001 97 2.5E+06 4.7E+04 0.019 57 & 58 I 32.000 24.000 1.000 22.000 69 1 .2E+06 2.4E+04 0.020 59 & 60 I 28.000 26.000 1.000 24.000 75 1 .3E+06 2.8E+I04 0.022 61 V 8.250 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.024 62 &63 II 15.750 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.024 64 VI 13.250 10.750 0.365 10.020 31 2.0E+05 4.7E+i03 0.024 66 V j8.250 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.024 67 III 12.750 12.750 0.375 12.000 38 2.4E+05 6.6E+03 0.028 68 II 16.000 18.000 0.500 17.000 53 4.5E+05 1.3E+04 0.029 69 -73 III 12.750 12.750 0.375 12.000 38 2.4E+05 6.6E+03 0.028 75 V 8.250 10.750 0.365 10.020 131 2.0E+'05 4.7E+03 0.024 76 V 8.250 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.024 77 & 78 II 15.750 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.024 79 II 12.000 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.024 80 II 16.000 18.000 0.500 17.000 53 4.5E+05 1.3E+'04 0.029 81 II 12.000 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.024 82 V 8.250 18.000 0.500 17.000 53 4.5E+05 1.3E+04 0.029 83 & 84 10.000 24.000 0.500 23.000 72 6.1E+05 2.4E+04 0.038 85 III 12.750 12.750 0.375 12.000 38 2.4E+05 6.6E+03 0.028 86 III 23.670 12.750 0.375 12.000 38 2.4E+05 6.6E+03 0.028 87 II 16.000 18.000 0.500 17.000 53 4.5E+05 1.3E+04 0.029 88 VII 15.250 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.024 90 VII 8.250 10.750 0.365 10.020 31 2.0E+05 4.7E+'03 0.024 91 -98 II 19.750 18.000 0.750 16.500 52 6.6E+05 1.3E+04 0.020 100 9.250 4.500 0.237 4.026 13 5.1E+04 8.3E+02 0.016 101 -104 I 20.550 18.000 0.750 16.500 52 6.6E+05 1.3E+04 0.020 Maximum PclFc = 0.038 V12, Page 11 of 17 SHEET FI-4 X6CNAI15 ATTACHMENT F1 Evaluation of 52 psig Pressure on UI Pipe Penetrations Shear[111= Type V or VII penetration; dimension B used instead of L Type V Penetration:
Per IX4DL4A0 14, there is no dimension L; use B instead Type VII Penetration:
If dimension L not provided on 1X4DL4A013 or 1X4DL4A014, use B instead PEN # IType L 0 t ID Lw Fs Ps JPs/Fs 1 -4 I 48.250 56.000 1.5001 53.0001167 2.9E+06 4.4E+05 0.153 5 VII 8.250 10.750 0.365 10.020 31.5 1.3E+05 1.4E+04 0.109 7- 10 I 18.000 18.000 0.500 17.000 53.4 3.1E+05 5.3E+04 0.171 11& 12 III 11.750 12.750 0.375 12.000 37.7 1.6E+05 2.4E+04 0.150 13 III 12.750 12.750 0.375 12.000 37.7 1.6E+05 2.7E+04 0.163 14 VII 15.250 10.750 0.365 10.020 31.5 1.3E+05 2.7E+04 0.202 15 II 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.208 16- 17 V 8.250 10.750 0.365 10.020 31.5 1.3E+05 1.4E+04 0.109 18-21 I 37.500 34.000 1.500 31.000 97.4 1.7E+06 2.1E+05 0.123 22 II 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.208 23 I 14.930 10.750 0.365 10.020 31.5 1.3E+05 2.6E+04 0.197 24 II 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.208 25 V 8.250 18.000 0.500 17.000 53.4 3.1E+05 2.4E+04 0.079 28 & 29 II 16.750 18.000 0.500 17.000 53.4 3.1E+05 4.9E+04 0.160 30 & 31 II 18.750 14.000 0.438 13.124 41.2 2.1E+05 4.3E+04 0.205 32 II 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.208 33 II 18.750 14.000 0.438 13.124 41.2 2.1E+05 4.3E+04 0.205 34 & 35 II 19.250 20.000 0.500 19.000 59.7 3.5E+05 6.3E+04 0.182 40 II 16.750 18.000 0.500 17.000 53.4 3.1E+05 4.9E+04 0.160 41 II 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.208 42 II 12.000 10.750 0.365 10.020 31.5 1.3E+05 2.1E+04 0.159 43-46 II 18.750 18.000 0.500 17.000 53.4 3.1E+05 5.5E+04 0.179 47 V 8.250 10.750 0.365 10.020 31.5 1.3E+05 1.4E+04 0.109 48 I 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.208 49 II 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.208 50 I 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.208 51 -55 I 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.208 V12, Page 12 of 17 SHEET Fl-5 X6CNA1 5 ATTACHMENT Fl Evaluation of 52 psig Pressure on UI Pipe Penetrations Shear I ... I=Type V or VII penetration; dimension B used instead of L Type V Penetration:
Per IX4DL4A0 14, there is no dimension L; use B instead Type VII Penetration:
If dimension L not provided on IX4DL4A013 or 1X4DL4A014, use B instead PEN # Type L D t ID Lw Fs Ps P 5/F 5 56 I 44.750 34.000 1.500 31.000 97.4 1.7E+06 2.5E+05 0.147 57 & 58 I 32.000 24.000 1.000 22.000 69.1 8.0E+05 1.3E+05 0.157 59 & 60 I 28.000 26.000 1.000 24.000 75.4 8.7E+05 1.2E+05 0.136 61 V o8.25'0 10.750 0.365 10.020 31.5 1.3E+05 1.4E+04 0.109 62 &63 II 15.750 10.750 0.365 10.020 31.5 i.3E+05 2.8E+04 0.208 64 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 0.109 67 III 12.750 12.750 0.375 12.000 37.7 1.6E+05 2.7E+04 0.163 68 II 16.000 18.000 0.500 17.000 53.4 3.1E+05 4.7E+04 0.152 69 -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 0.109 77 & 78 II 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.208 79 II 12.000 10.750 0.365 10.020 31.5 1.3E+05 2.IE+04 0.159 80 II 16.000 18.000 0.500 17.000 53.4 3.IE+05 4.7E+04 0.152 81 II 12.000 10.750 0.365 10.020 31.5 1.3E+05 2.1E+04 0.159 182 ....8.250 18.000 0.500 17.000 53.4 3.1E+05 2.4E+04 0.079 83 & 84 10.000 24.000 0.500 23.000 72.3 4.2E+05 3.9E+04 0.094 85 III 12.750 12.750 0.375 12.000 37.7 1.6E+05 2.7E+04 0.163 86 III 23.670 12.750 0.375 12.000 37.7 1.6E+05 4.9E+-04 0.302 87 II 16.000 18.000 0.500 17.000 53.4 3.IE+05 4.7E+04 0.152 88: 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 0.109 91 -98 II 19.750 18.000 0.750 16.500 51.8 4.5E+05 5.8E+04 0.129 100 9.250 4.500 0.237 4.026 12.6 3.5E+04 6.8E+03 0.196 101,-104 I 20.550 18.000 0.750 16.500 51.8 4.5E+05 6.0E+04 0.134 Maximum PslFs = 0.302 V12, Page 13 of 17 X6CNA15 ATTACHMENT F2 SHEET F2-1 Evaluation 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 ftemp ac-norn = 20,000 psi = Nominal allowable comprssive stress ftermp = 0.85 = Reduction due to increased temPerature O-c = 17,000 psi = Allowable compressive stress as = Allowable shear stress (psi)a's = a's-nom X ftemp a'c-nomn = ,I13,600 psi = Nominal allowable comprssive stress ftemp = 0.85 = RedUction due to increased temperature O's = 11,560 psi = Allowable shear stress Fc = ac x Lw x t = Allowable Compressive Load (Ibf)Fs = as x Lw x t = Allowable Shear Load (Ibf)Pctmnt = 52 psig = Containment Pressure Pc = 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 17 SHEET F2-2 X6CNA15S ATTACHMENT F2 Evaluation of 52 psig Pressure on U2 Pipe Penetrations Compression I I= Type V or VII penetration; dimension B used instead of L iType V Penetration:
Per 2X4DL4A0 14, there is no dimension L; use B instead Type VII Penetration:
If dimension L not provided on 2X4DL4A013 or 2X4DL4A014, use B instead PEN # Type L D t ID Lw FcPc PclFc 1 -4 I 48.25 56.000 1.500 53.000 167 4,2E+06 1.3E+05 0.030 5 VII 8.250 10.750 0.365 10.020 31 2,0E+05 4,7E+03 0.024 7- 10 I 18.000 18.000 0.500 17.000 53 4.5E+05 1,3E+04 0.029 11& 12 III 11.750 12.750 0.375 12.000 38 2,4E+05 6,6E+03 0.028 13 III 12.750 12.750 0.375 12.000 38 2.4E+05 6.6E+03 0.028 14 VII 15.250 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.024 15 II 15.750 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.024 16 -17 V 8.250 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.024 18-21 I 37.500 34.000 1.500 31.000 97 2.5E+06 4,7E+04 0.019 22 II 15.750 10.750 0.365 10.020 31 2,0E+05 4,7E+03 0.024 23 I 14.930 10.750 0.365 10.020 31 2.0E+05 4,7E+03 0.024 24 II 15.750 10.750 0.365 10.020 31 2,0E+05 4,7E+03 0.024 25 ,V 8.250 18.000 0.500 17.000 53 4.5E+05 1,3E+04 0.029 28 & 29 II 16.750 18.000 0.500 17.000 53 4.5E+05 1,3E+04 0.029 30 & 31 II 18.750 14.000 0.438 13.124 41 3.1E+05 8,0E+03 0.026 32 II 15.750 10.750 0.365 10.020 31 2,0E+05 4,7E+03 0.024 33 II 18.750 14.000 0.438 13.124 41 3,1E+05 8,0E+03 0.026 34 & 35 II 19.250 20.000 0.500 19.000 60 5.IE+05 1,6E+04 0.032 40 II 16.750 18.000 0.500 17.000 53 4,5E+05 1,3E+04 0.029 41 II 15.750 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.024 42 II 12.000 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.024 43 -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 0.024 48 I 15.750 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.024 49 II 15.750 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.024 50 I 15.750 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.024 51 -55 I 15.750 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.024 V12, Page 15 of 17 SHEET F2-3 X6CNA15S ATTACHMENT F2 Evaluation of 52 psig Pressure on U2 Pipe Penetrations Compression
~=TpeorI=
pentraIon dimension Bused instead of L Type V Penetration:
Per 2X4DL4A0 14, there is no dimension L; use B instead Type VII Penetration:
If dimension L not provided on 2X4DL4A013 or 2X4DL4A014, use B instead PEN # Type L D t ID LFcPc Pc/Fc 56 I 44.750 34.000 1.500 31.000 97 2.5E+06 4.7E+04 0.019 57 & 58 I 32.000 24.000 1.000 22.000 69 1.2E+06 2.4E+04 0.020 59 & 60 1 28.000 26.000 1.000 24.000 75 1 .3E+06 2.8E+04 0.022 61 V 8.250 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.024 62 &63 II 15.750 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.024 64 VI 13.250 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.024 66 V 8.250 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0,024 67 III 12,750 12,750 0.375 12,000 38 2,4E+05 6.6E+03 0,028 68 II 16.000 18.000 0.500 17.000 53 4.5E+05 1.3E+04 0.029 69-73 III 12.750 12.750 0.375 12.000 38 2.4E+05 6.6E+03 0.028 75 V 8.250 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.024 76 V 8.250 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.024 77 &78 II 15.750 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.024 79 II 12,000 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.024 80 II 16.000 18.000 0.500 17.000 53 4.5E+05 1.3E+04 0.029 81 II 12.000 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.024 82 V 8.250 18.000 0.500 17.000 53 4.5E+05 1.3E+04 0.029 83 &84 10.000 24.000 0.500 23.000 72 6.IE+05 2.4E+04 0.038 85 III 12.750 12.750 0.375 12.000 38 2.4E+05 6.6E+03 0.028 86 III 23.670 12.750 0.375 12.000 38 2.4E+05 6.6E+03 0.028 87 II 16.000 18.000 0.500 17.000 53 4.5E+05 1.3E+04 0.029 88 VII 15.250 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.024 90 VII 8.250 10.750 0.365 10.020 31 2.0E+05 4.7E+03 0.024 91 -98 II 19.750 18.000 0.750 16.500 52 6.6E+05 1.3E+04 0.020 100 9.250 4.500 0.237 4.026 13 5.1E+04 8.3E+02 0.016 101 -104 I 20.550 18.000 0.750 16.500 52 6.6E+05 1.3E+04 0.020 Maximum PdIFc = 0.038 V12, Page 16 of 17 SHEET F2-4 X6CNA15S ATTACHMENT F2 Evaluation of 52 psig Pressure on U2 Pipe Penetrations Shear Type V orVillpntain dimension B used instead of L Type V Penetration:
Per 2X4DL4A014, there is no dimension L; use B instead Type VII Penetration:
If dimension L not provided on 2X4DL4A013 or 2X4DL4A014, use B instead PEN # Type L D t ID Lw Fs Ps PslFs 1 -4 I 48.250 56.000 1.500 53.000 167 2.9E+06 4.4E+05 0.153 5 VII 8.250 10.750 0.365 10.020 31.5 1,3E+05 1.4E+04 0.109 7-10 I 18.000 18.000 0.500 17.000 53.4 3.1E+05 5.3E+04 0.171 11& 12 III 11.750 12.750 0.375 12.000 37.7 1.6E+05 2.4E+04 0.150 13 III 12.750 12.750 0.375 12.000 37.7 1.6E+05 2.7E+04 0.163 14 VII 15.250 10.750 0.365 10.020 31.5 1.3E+05 2.7E+04 0.202 15 II 15.750 10.750 !0.365 10.020 31.5 1.3E+05 2.8E+04 0.208 16-17 V 8.250 10.750 0.365 10.020 31.5 1.3E+05 1.4E+04 0.109 18-21 1 37.500 34.000 1.500 31.000 97.4 1.7E+06 2.IE+05 0.123 22 II 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.208 23 I 14.930 10.750 0.365 10.020 31.5 1.3E+05 2.6E+04 0.197 24 II 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.208 25 V 8.250 18.000 0.500 17.000 53.4 3.1E+05 2.4E+04 0.079 28 &29 II 16.750 18.000 0.500 17.000 53.4 3.1E+05 4.9E+04 0.160 30 & 31 II 18.750 14.000 0.438 13.124 41.2 2.1E+05 4.3E+04 0.205 32 II 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.208 33 II 18.750 14.000 0.438 13.124 41.2 2.IE+05 4.3E+04 0.205 34 & 35 II 19.250 20.000 0.500 19.000 59.7 3.5E+05 6.3E+04 0.182 40 II 16.750 18.000 0.500 17.000 53.4 3.IE+05 4.9E+04 0.160 41 II. 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.208 42 II 12.000 10.750 0.365 10.020 31.5 1.3E+05 2.1E+04 0.159 43-46 II 18.750 18.000 0.500 17.000 53.4 3.1E+05 5.5E+04 0.179 47 V 8.250 10.750 0.365 10.020 31.5 1.3E+05 1.4E+04 0.109 48 I 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.208 49 II 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.208 50 I 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.208 51-55 I 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.208 V12, Page 17 of 17 SHEET F2-5 X6CNA1 5 ATTACHMENT F2 Evaluation of 52 psi 9 Pressure on U2 Pipe Penetrations Shear[111= Type V or VII penetration; dimension B used instead of L Type V Penetration:
Per 2X4DL4A014, there is no dimension L; use B instead Type VII Penetration:
If dimension L not provided on 2X4DL4A013 or 2X4DL4A014, use B instead PEN # Type L D t ID L FsPs PslFs 56 I 44.750 34.000 1.500 31.000 97.4 1.7E+06 2.5E+05 0.147 57 & 58 I 32.000 24.000 1.000 22.000 69.1 8.0E+05 1.3E+05 0,157 59 & 60 I 28.000 26.000 1.000 24.000 75.4 8.7E+05 1.2E+05 0.136 61 V 8.250 10.750 0.365 10.020 31.5 1.3E+05 1.4E+04 0.109 62 &63 II 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.208 64 VI 13.250 10.750 0.365 10.020 31.5 1.3E+05 2.3E+04 0.175 66 V 8.250 10.750 0.365 10.020 31.5 1.3E+05 1.4E+04 0.109 67 III 12.750 12.750 0.375 12.000 37.7 1.6E+05 2.7E+04 0.163 68 II 16.000 18.000 0.500 17.000 53.4 3.1E+05 4.7E+04 0.152 69-73 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 76 V 8.250, 10.750 0.365 10.020 31.5 1.3E+05 1.4E+04 0.109 77 &78 II 15.750 10.750 0.365 10.020 31.5 1.3E+05 2.8E+04 0.208 79 II 12.000 10.750 0.365 10.020 31.5 1.3E+05 2.1E+04 0.159 80 II 16.000 18.000 0.500 17.000 53.4 3.1E+05 4.7E+04 0.152 81 II 12.000 10.750 0.365 10.020 31.5 1.3E+05 2.1E+04 0.159 82 V 8.250 18.000 0.500 17.000 53.4 3.1E+05 2.4E+04 0.079 83 & 84 10.000 24.000 0.500 23.000 72.3 4.2E+05 3.9E+04 0.094 85 III 12.750 12.750 0.375 12.000 37.7 1.6E+05 2.7E+04 0,163 86 III 23.670 12.750 0.375 12.000 37.7 1.6E=+05 4.9E=+04 0.302 87 II 16.000 18.000 0.500 17.000 53.4 3.1E+05 4.7E+04 0.152 88 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 0.109 91 -98 II 19.750 18.000 0.750 16.500 51.8 4.5E+05 5.8E+04 0.129 100 9.250 4.500 0.237 4.026 12.6 3.5E+04 6.8E+03 0.196 101 -104 I 20.550 18.000 0.750 16.500 51.8 4.5E+05 6.0E+-04 0.134 Maximum PslFs 0.302 I-+/- 4.4.4 .J~XH~_ ~ Th.7 r~.j Vi3 Page 1 of 1 T<-H........ .....--I ALVI W. VOCTHE P1AJ hC.;a i i a i V1 4 Page 1 of 2 provided 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 located inside battery rooms.3.2 POWER GENERATION DESIGN BASES None 3.3 MAJOR COMPONENT DESIGN BASES 3.3.1 The Class IE dc 125-V dc system and all equipment are classified as Seismic Category 1 and 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 voltage range of 105 to 14OV dc.The 25 kVA inverters provide power to RIHR isolation valves. The breakers for these valves are located in "Trip" position during normal plant operation; they do not operate during the duty cycle time of the battery, or for any LOCA/LOOP or SBO coping scenario.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 required loads for a loss-of-coolant accident, loss of offsite power, or main steam line break for a duration 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 battery replacement 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 6VR1 6 VER 13 V14 Page 2 of 2 3.3.3.7 A dc ammeter, dc voltmeter, dc overvoltage relay, ac power "on" light, and ac undervoltage relay shall be provided on each charger. The relays shall alarm in the control 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 to recharge a discharged battery in a shorter amount of time (if allowed by the manufacturer).
3.3.3.9 The battery charger shall prevent the charger from becoming a load on the battery due to a 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 the charger from the ac source.3.3.3.11 dc breakers shall be provided to protect the battery and charger from internal charger faults 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 vital instrumentation 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 shall also 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-phase inverter 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 105 to 140 V dc. The dc feeder cables shall be designed to maintain a minimum of 105 V dc during 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 to 140 V dc. The dc feeder cables to the RHR isolation valve inverters shall be designed to maintain a minimum of 105 V dc at the RH-R isolation valve inverters.
3.3.5 125-V dc MCCs 3.3.5.1 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 8VR1 8 VER 13 X6CNA1 5 Attachment L
C2I.l~at~inn fnr I:W.I I1 SHEET L-6 v15 Page 1 of 3 CALC NO. SNC024-CALC-004 VEGP DETERMINATION OF EMERGENCY ACTION LEVEL REV. 0 I0 E N E R C 0 N FOR INITIATING CONDITION E-PAGE NO. Page 6of 8 5.0 Design Inputs 1. The contact dose rates from the HI-STORM 100 and HI-TRAC 125 cask system technical specification
[2, Table 6.2-3] are provided below in Table 5-1. These source 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-STORM 100 and HI-TRAC 125 Number of Technical Specification LoatonjMeasurements j Limit (mrem/hr)HI-TRAC 1.25 __________
Side -Mid -height 4 472.7 Top 1 4 j102.4 HI-STORM 100 ____ _____Side -60 inches below mid-height 4 87 Side -Mid -height 4 88.9 Side -60 inches above mid-height 4 54.8 Top -Center of lid 1 24.5 Top -Radially centered 4 29.2 Inlet duct 4 178.8 Outlet duct 4 64.5 X6CNA1 5 Attachment L ENERCON Calculation for E-HU1 SHEET L-7 v15 Page 2 of 3 CALC NO. SNC024-CALC-004 VEGP DETERMINATION OF ____________
EMERGENCY ACTION LEVEL REV. 0 0 E N E R C 0 N FOR INITIATING CONDITION E- -__________
PAGE NO. Page 7of 8 6.0 Methodology The "on-contact" dose rates from the technical specification for the HI STORM-i100 cask system are scaled by a factor of 2, as specified in NEI 99-01 Rev. 6 [1], for use in initiating condition E-HUI.
X6CNA1 5 Attachment L FNFRf.ON ('nlrmiitinn fnr F-HI- 11 SHEET L-8V1 Page 3 of 3 CALC NO. SNC024-CALC-004 VEGP DETERMINATION OF -____________
EMERGENCY ACTION LEVEL REV. 0 0 E N E R C 0 N FOR INITIATING CONDITION E-PAGE NO. Page 8of 8 7.0 Calculations The dose rates in Table 5-1 are multiplied by 2 in order to calculate the EAL dose rate limits. These calculations are presented below in Table 7-1.Table 7-1 Dose Rate Scaling Calculations for EAL Limits Technical LoainSpecification Scaling Calculated Value EAL LoainLimit Factor (mrem/hr) (mrem/hr)_________________________ (mrem/hr) j____ _______________
HI-TRAC 125___ _____Side -Mid -height [ 472.7 2 1 945.4 [ 950 Top 102.4 2 j 204.8 [ 200 HI-STORM 100 Side -60 inches below mid-height 87 2 174 170 Side -Mid -height 88.9 2 177.8 180 Side -60 inches above mid-height 54.8 2 109.6 110 Top -Center of lid 24.5 2 49 50 Top -Radially centered 29.2 2 58.4 60 Inlet duct 178.8 2 357.6 360 Outlet duct 64.5 2 129 130 8.0 Computer Software Microsoft WORD 2013 is used in this calculation for basic multiplication.
V1 6 Page 1 of 4 Approved By J. B. Stanley Effective Date 7/25/12~Voqgtle Electric Generating F-0 CRITICAL SAFETY FUNCTION Si F- 0.2 CORE COOLING PatProcedure version Pat19200-C 24.2 Page Number'ATUS TREES5ol Sheet 1 of 1-4 GO TO 11221-V~GO TO 19221-C VLI5 FULL NO GE GREATER tHAN 41% ,E8S* .GO TO 19222-C 11.FI GO TO 19222Z-C tVUI FULL N iGE GREATER THAN 41% YE.GO TO* 19223-V~GO TO rYNAMIC Ha 9,Z-INGE N ho -4 RCP h-3 RCP h-2RCP E-1IRCF* 19223-C GI CSA T v Printed February 2, 2016 at 16:09 Vi16 Page 2 of 4 Approved By Procedure Veso J. B. Stanley Vogtle Electric Generating P..lantng 19200-C Veso24.2 Effective Date Page Number 7/25/12 F-0 CRITICAL SAFETY FUNCTION STATUS TREES6of1 Sheet 1 of 1 F- 0.3 HEAT SINK I KTOTAL AVAILABLE FEEDWATER FLOW TO SGs GREATER iTHAN 570 GPM NO YES NARROW RANGE LEVEL IN AT LEAST rONE SG GREATER THAN 10% (32%)GO TO 1 9231 -C GO TO 19232-C GO TO 1 9233-C NO PRESSURE IN ALL SGs LESS THAN 1240 PSIG YE LESS THAN 82% YES NO PRESSURE IN ALL SGs LESS THAN 1180 PSIG YES GO TO 1 9234-C GO TO 19235-C NARROW RANGE NO LEVEL IN ALL SGs GREATER THAN 10% (32%) YES L SAT Printed February 2, 2016 at 16:09 V1 6 Page 3 of 4 Approved By Procedure Version J. B. Stanley Vogtle Electric Generating Plant 19200-C 24.2 Effective Date Page Number 7/25/12 F-0 CRITICAL SAFETY FUNCTION STATUS TREES7of1 Sheet 1 of 1 F- 0.4 INTEGRITY GO TO 19241-C---- .) GO TO I 19241-C ALL RCS WR COLD LEG TEMPERATURES GREATER THAN 295 0 F~QGO TO 19242-C~*CSF SAT TEMPERATURE DECREASE IN ALL NO RCS COLD LEGS LESS100 0 F IN THE ILAST iYES!60 MINUTES JGOTO I,11 19241-C ALL RCS WR COLD LEG NO TEMPERATURES GREATER THAN 265°F YES RCS PRESSURE LESS NO THAN COLD OVERPRESSURE LIMIT 465 PSIG YES ,I RCS WR COLD LEG NO TEMPERATURE GREATER THAN 220 0 F YES LL* (I'\ GO TO" "" 19242-C O CSF SAT-*CSF SAT Printed February 2, 2016 at 16:09 Vi16 Page 4 of 4 Approved By .....Procedure Version J. B. StanI~yVogtle Electric Generating Plant '19200-C 2.Effe tiv Datan ey P... ..age... :.. .... .... ... ...... Numbe Efeciv DteF-0 CRITICAL SAFETY FUNCTION STATUS TREES 9ag ofmb1r 7/25/129of1 Sheet 1 of 1 F- 0.5 CONTAINMENT GO TO 19251-C* )GO TO i 19251-C N.E N o P GO TO p 19252-C I CONTAINMENT BUMP LEVEL LESS THAN 196 iNCHESGO TO.......19253.C Ib CSF SAT Printed February 2, 2016 at 16:09 V1 7 Page 1 of 3 Approved By p, 4Pocdr Version W. L. Burmeister
... Vogtle Electric Generating Plant IS55039-c~rcdr 3.2 Effective Date P~age Number 0510712013 ISEISMIC MONITORING INSTRUMENTATION SYSTEM I 6 of 9 4.2 NORMAL OPERATION NONE 4.3 NON-PERIODIC OPERATIONS NOTE This subsection shall be initiated by 50022-C "Seismic Event Plan" or Supervisor's direction.
1 4.3.1 Retrieving Seismic Data NOTE SKey Number 1-OP3-10 is in C&T. LI a. Verify the Event alarm on the Condor Control Unit screen is RED. El b. Obtain the event charts and graphs from each recorder.
El 4.3.2 Retrieving Seismic Data At ETNA (River Intake)a. Hook up a laptop to the uplink cable. El b. Power on computer, THEN click on the ALTUS Quick Talk icon. El c. In the ALTUS Status window, verify the alarm has been triggered.
El d. 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 recorded event El (3) Click on the Retrieve File button El (4) Download the file to a floppy disk El Printed February 15, 2016 at 16:59 V17 Page 2 of 3 Seismic Event Plan 50022-C VOGTLE Version 14.0 Unit C Page 7of 24 4.0 INSTRUCTION 4.1 IDENTIFICATION OF SEISMIC EVENT 1. The following indicators are available for determining whether or not seismic event 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 has occurred; (Must be called when any of the above indications of an earthquake is received) and to initiate this procedure if earthquake is confirmed.
- 2. IF an earthquake is sensibly detected by control room personnel, initiate this procedure.
- 3. IF AOP 18036-C, Seismic Event, is in effect, then initiate this procedure.
Printed February 15, 2016 at 17:02 V1 7 Page 3 of 3 Seismic Event 18036-C VOGTLE Version 11 Unit C Page 4of 9 PURPOSE The purpose of this procedure is to provide operator response following a seismic event and to initiate 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 VI18 Page 1 of 5 Southern Nuclear Design Calculation IPlant: Vogte Unit: 1&2 Icalculation Number: X6CNAI4 Isheet:;46 Miscellaneous Design Inputs 21. 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, which are 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 r Isotope Average Gamma Energy (MeV)
Reference:
Brookhaven National Laboratory National Nuclear Data Center decay data (http://www.orau .or qlptp/PTP%20Libraryllibrary/DOE/bnl/nu clidedata/table.htm)
Copies of web pages in Attachment C8 1-131 0.382 1-132 2.20 1-133 0.607 I-134 2.50 1-135 1.55 Co-580.975 Co-60 2.51 Cs-134 1.55 Cs-136 2.12 Cs-i137 0.565 Cs-138 2.31 Cs-138 2.31 Vi18 Page 2 of 5 Southern Nuclear Design Calculation SPlant: Vogtle Unit: 1&2 ICalculation Number: X6CNA14 ISheet: 61 Recognition Category S: System Malfunctions Notice of Unusual Event SU4: Fuel Clad Degradation.
Operating Mode Applicability:
Power Operation (Mode 1)Startup (Mode 2)Hot Standby (Mode 3)Hot Shutdown (Mode 4)1 OR2 Emergency Action Levels: SU4 EALI: CVCS Letdown radiation monitor RE-48000 reading greater than 5 pCi/cc indicating 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/g Per section B.3.4.16, page B3.4.16-2 of VEGP Tech Spec Bases, noble gas activity in the reactor coolant assumes 1% failed fuel, which closely equals the 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 13 of X6AZ01 A, the principle isotopes detected by RE-48000 are 1-131, 1-133, Co-58, Co-60, Cs-134, and Cs-137.However, per Section B-12-3-2 and Figure B-12-2 of 1X6AZ01-10004 & 2X6AZ01-10004, RE-48000 will detect gammas of energies down to-0.1 MeV.S t 1 __ _ "__I-, -.. I _ _ _ _i c -i ..L =t 4 ..II; P.m IENKR4Y It ,VgL= t.VI Figure B-12-2 Thus the other I, Co, and Cs isotopes listed in FSAR Table 11.1-2 should be included if their average decay gamma energies exceed 0.1 MeV.
V1 8 Page 3 of 5 Southern Nuclear Design Calculation iPlant: Vogtle U nit: 1&2 ICalculation Number: X6CNAI4 Sheet: 62 Per LTR-CRA-06-179 attached to WEC-SNC letter GP-18006, the pre-MURPU coolant activities may be adjusted upward 2% to account for the increase in core thermal power from 3565 MWt to 3636 MWt. Thus, the Co and Cs MURPU 1% defect activity are equal to their pre-MURPU 1% Defect activities multiplied by 1.02.The Co and Cs activities corresponding to the 1.0 DE 1-131 Tech Spec limit are the products of their MURPU 1% defect activities and the ratio 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 the CVCS 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% Defect Isotope Coolant Coolant Coolant Activity Activity Activity I-131 0.74 2.91 ______I-132 0.75 2.96 ______I-133 1.41 5.56 1-134 0.18 0.69 ______I-135 0.69 2.72 ______Co-58 3.89E-03 1 .53E-02 1 .50E-02 Co-60 4.93E-04 1 .94E-03 1 .90E-03 Cs-134 5.97E-01 2.35 2.3 Cs-I136 7.52E-01 2.96 2.9 Cs-137 3.89E-01 1.53 1.5 Total = 5.5 21.7 ptCi/g Total = 5.5 21.5 i.LCi/cc CVCS Letdown Density =0.99 g/cc SGiven 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 8 Southern Nuclear Design Calculation Page 4 of 5 SPlant: Vogtle Unit: 1&2 Calculation Number: X6CNA14 Sheet: C5-1 Attachment C5 -VEGP 1&2 CVCS Letdown Radiation Monitor (RE-48000)
Readings U 1.... ... .a IIII-~' I~~-:-~a..
V1 8 Southern Nuclear Design Calculation Page 5 of 5 SPlant: Vogtle Unit: 1&2 Calculation Number: X6CNA14 Sheet: C5-2 Attachment C5 -VEGP 1&2 CVCS Letdown Radiation Monitor (RE-48000)
Readings*]W11 o=I~IMY Iin'~vu l~u~r ~.~tImP~.ii I~' ~'~I WE ~'~~jL 11U WOWW4m~ ~
V19 Page 1 of 3 RCS Specific Activity 3.4.16 3.4 REACTOR COOLANT SYSTEM (RCS)3.4.16 RCS Specific Activity LCO 3.4.16 APPLICABILITY:
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--------------------------
INlJ LCO 3.0.4c is applicable.
I------------------
---CONDITION REQUIRED ACTION COMPLETION TIME A. DOSE EQUIVALENT A.1 Verify DOSE Once per4 hours I-131 > 1.0 p.Ci/gm. EQUIVALENT I-131 within the acceptable region of Figure 3.4.16-1.AND A.2 Restore DOSE 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> EQUIVALENT I-131 to within limit.B. Gross specific activity of B.1 Perform SR 3.4.16.2.
4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> the reactor coolant not within limit. AND 8.2 Be in MODE 3 with 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Tavg < 500°F.(continued)
Vogtle Units 1 and 2 3.4.16-1 Amendment No. 137 (Unit 1)Amendment No. 116 (Unit 2)
V1 9 Page 2 of 3 RCS Specific Activity 3.4.16 ACTIONS (continued)
________________
__________
CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and C.1 Be in MODE 3 with 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Tavg < 500°F.Time of Condition A not met.O_.R DOSE EQUIVALENT 1-131 in the unacceptable region of Figure 3.4.16-1.SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.16.1 Verify reactor coolant gross specific In accordance with activity_
100/I. !iCi/gm. the Surveillance Frequency Control Program SR 3.4.16.2 ---- --NOTE- --- --Only required to be performed in MODE 1.Verify reactor coolant DOSE EQUIVALENT I-131 In accordance with specific activity < 1.0 ,.tCi/gm, the Surveillance Frequency Control Program AND Between 2 and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after a THERMAL POWER change of _> 15% RTP within 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 2 3.4.16-2 Amendment No. 158 (Unit 1)Amendment No. 140 (Unit 2)
V1 9 Page 3 of 3 RCS Specific Activity 3.4.16 250 I U-I.200 150 100 50 PERCENT OF RATED THERMAL POWER FIGURE 3.4.16-1 REACTOR COOLANT DOSE EQUIVALENT I-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACITVITY >1 mCi/gram DOSE EQUIVALENT 1-131 Vogtle Units 1 and 2 3.4.16-4 Amendment No. 96 (Unit 1)Amendment No. 74 (Unit 2)
V20 Page 1 of 1 RCS Operational LEAKAGE 3.4.13 3.4 REACTOR COOLANT SYSTEM (RCS)3.4.13 RCS Operational LEAKAGE LCO 3.4.13 RCS operational LEAKAGE shall be limited to: a. No pressure boundary LEAKAGE;b. I1 gpm unidentified LEAKAGE;c I1 p dniidLAAE n d. 150 galosper idaytprimdLAryGE toscndar EKG hog n one steam generator (SG).APPLICABILITY:
MODES 1, 2, 3, and 4.ACTIONS__________________
___CONDITION REQUIRED ACTION COMPLETION TIME A. RCS operational A.1I Reduce LEAKAGE to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> LEAKAGE not within within limits.limits for reasons other than pressure boundary LEAKAGE or primary to secondary LEAKAGE.B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition A not AND met.B.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> O__R Pressure boundary LEAKAGE exists.OR Primary to secondary LEAKAGE not within limit.Vogtle Units 1 and 2 3.4.13-1 Amendment No. 144 (Unit 1)Amendment No. 124 (Unit 2)
V21I Page 1 of 2 Approved ByPoede Vrin J. B. Stanley Vogtle Electric Generating Plant 19200cdr 24.2i~Effective Date -0CTIASAEYFNTOSAUSRES Page Number 7/25/12 F 0 C I CA SA E Y F N T O ST T S R ES9 of 11 Sheet 1 of 1 F- 0.5 CONTAINMENT GOaTO 19251-C.-'- PRESURELS u ==s4 j TANji sl j I* a O T 19251-(;I J i .AT LEAST ONE SCONTAINMENT SSPRAY PUMP SRUNNING NO YES*ego GO TO* il ) 19261-C I GO TO 19252-C b/T> GO TO: ......1 9 2 6 3 -C CSP SAT ,-r'nneu rebruary iZUll at 14:zz V21 Page 2 of 2 S0uthern Nuclear Operating Company A~rlN Plant: VEGP ! "X6CNA1 5 ISUHda Unit: 1&2 Title: NEI 99-01 Rev 6 EAL Calculations SHEET 43 If Containment pressure (PCTMT) exceeds the static head (AH) dlue to the difference between the Transfer 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), the Transfer 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/ft 3) x g1 (ft/sec2) x 1 ft 2 go (Ibm -ft)/(lIbf-sec
- 2) 144 in 2 Pctmt (psig) > 30 ft x 61.55 Ibm x 32.2 (ft/sec2) x 1 ft 2 t 3 32.2 (Ibm-ft)/(Ibf-sec
- 2) 144 in 2 (Design Input #25)Petmi > '43 psig Pressure > 52 psig WITH Tech Spec containment integrity intact NMP-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. 6.1.1 states "Perform required visual examinations and leakage rate testing except for containment air lock testing, in accordance with the Containment Leakage Rate Testing Program." Tech Spec section 5. 5.17 describes the Containment Leakage Rate Testing Program. Per Tech Spec Bases B3. 6.1, the Containment is designed to contain radioactive material that may be released from 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, these penetrations 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; see Attachment CS) and Ron Cowen (Westinghouse Site Services Manager; see Attachment C6).