ML14164A308

Project, Revision to Unit 1 and Unit 2 Emergency Action Levels, Evaluation of the Proposed Changes License Amendment Request (LAR) for Revision of EAL Scheme to NEI 99-01 Revision 6. Part 5 of 7

ML14164A308
Person / Time
Site:	South Texas
Issue date:	05/15/2014
From:	South Texas
To:	Office of Nuclear Reactor Regulation
Shared Package
ML14164A341	List: ML14164A300 ML14164A303 ML14164A305 ML14164A308 ML14164A311 ML14164A314 ML14164A318 NOC-AE-14003087, Revision to Unit 1 and Unit 2 Emergency Action Levels, Evaluation of the Proposed Changes License Amendment Request (LAR) for Revision of EAL Scheme to NEI 99-01 Revision 6. Part 7 of 7
References
NOC-AE-14003087
Download: ML14164A308 (361)

Text

OPOP03-ZG-0007 I Rev. 71 Page 197 of216 Plant Cooldown I Addendum 19 Controlling RCS Inventory at or above Elv 39 ft. 4.9 in. Page 3 of 10 INITIALS CAUTION" Do NOT Reduce RCS level Below Ely 39 ft. 4.9 in. using this Addendum.* As pressurizer level lowers to 10% AND before the PZR level goes off scale low, COMPARE Pressurizer level with RC-LG-3662 "RCS LEVEL SIGHTGLASS (SLINKY)" level indication.

3.0 PERFORM

the following to lower RCS inventory:

3.1 The following instrumentation SHALL be operable prior to RCS Draindown to 39' 4.9": 3.1.1 RHR heat exchanger inlet and outlet temperature with indication on QDPS OR chart recorder for all operable trains.3.1.2 RHR pump flow with indication on QDPS for all operable trains.3.1.3 RHR pump motor current indication (amps) for all operable trains.3.1.4 "RHR PUMP CURRENT LO" Annunciators on Lampbox I M02 for the operable RHR pumps is NOT removed from service.3.1.5 RCS level sightglass has been walked down in last twelve hours and satisfies requirements.

3.1.6 Both trains of RVWL are operable with at least two QDPS displays available.

3.1.7 Core Exit Thermocouples, five per train, two trains.3.2 ENSURE Adequate capacity is available in radwaste to receive volume drained from RCS. REFER TO Addendum 3, Determination of RCS Volume to be Drained.3.3 A dedicated, reliable communication line, headphones being the preferred method, is established between Control Room personnel and RCB sightglass watch.This procedure, when completed, SHALL be retained.

I OPOP03-ZG-0007 Rev. 71 Page 198 of216 j Plant Cooldown Addendum 19 Controlling RCS Inventory at or above Ely 39 ft. 4.9 in. Pa-e 4 of 10 j INITIALS NOTE* Venting and draining operations should be coordinated with the Radwaste Operator and Health Physics." RCS level changes SHALL be made slowly in a controlled manner to minimize effects on reactor vessel level indications.

The RCS level sightglass SHALL be continuously monitored during all draining and refill operations." One of the following pressurizer vent paths SHALL be established prior to draining the pressurizer:

• The Pressurizer spray line vent valves RC-0502 and RC-0503 OPEN to atmosphere.(Preferred Method)" A minimum of one Pressurizer Code Safety Valve REMOVED.* In addition to the specified temperature limits, the intent is to maintain RCS temperature as low as allowed by existing plant conditions, core cooling capabilities or other limiting criteria to maximize margin to core boiling:* (Mode 5) MAINTAIN RCS core exit temperature (RHR HX inlet temp when CETs NOT available) less than 140'F.3.4 IF PZR level wil be lowered below 10% Cold Calibrated level, THEN ENSURE RCS level sightglass is in service.This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 199 of 216 Plant Cooldown I Addendum 19 Controlling RCS Inventory at or above Ely 39 ft. 4.9 in. Page 5 of 10 INITIALS NOTE WHEN RC-LG-3662 "RCS LEVEL SIGHTGLASS (SLINKY)" is in service, THEN ENSURE "0POP07-RC-0001, RC Vent Rig/Sightglass Installation and Removal", LINEUP 1,"RC-LG-3662 "RCS LEVEL SIGHTGLASS (SLINKY)" in Service Lineup" PERFORMED daily and DOCUMENTED in a temporary log.3.5 COMMENCE temporary logging of LG-3662 sightglass valves once per shift (Ref. Procedure Step 3.37).CAUTION Prior to draining the pressurizer, a vent path SHALL be established to prevent drawing a vacuum.3.6 IF lowering pressurizer level to below 10% Cold Calibrated level (55 ft 6 inch elevation), THEN PERFORM the following:

3.6.1 ENSURE

RCS level sightglass is being monitored.

3.6.2 ENSURE

pressurizer vent path established.

3.6.3 DOCUMENT

Vent Path.3.6.4 OPEN PZR Spray Valves RC-PCV-655B and RC-PCV-655C.

3.6.5 DETERMINE

RCS volume to be drained using Addendum 3, Detennination of RCS Volume to be Drained.3.6.6 VERIFY "DIVERT LCV-01 12A" in the AUTO position.

{CP004}3.6.7 Manually RAISE letdown flow using "PRESS CONT PCV-0 135".{CP004}3.6.8 COMPARE Pressurizer level with RCS level sightglass level indication (Should agree within 6 inches. 1% of Pressurizer Cold Calibrated level equals 4 inches).This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 200 of216 Plant Cooldown Addendum 19 Controlling RCS Inventory at or above Elv 39 ft. 4.9 in. Page 6 of 10 INITIALS CAUTION IF RVWL system indicates less than 100%, THEN the draindown SHALL be stopped and the level difference between RVWL and RCS level sightglass investigated.

3.6.9 IF pressurizer Cold Calibrated level AND RCS level sightglass do NOT agree within 6 inches, THEN STOP the drain and REFER TO Addendum 15, Rx Head Venting and RCS level Instruments Disagreements Evaluation Guideline.

3.6.10 IF pressurizer level will be maintained, THEN REFER TO Step 3.22 of this Addednum.CAUTION IF during any RCS draining process, fluctuations are observed in RHR pump flow, amps, OR discharge pressure, THEN any RCS drain in progress SHALL be stopped to allow RCS water level to stabilize and any RCS water level recovery SHALL be initiated as necessary to ensure RHR system operation.

3.7 PLACE

Reactor Vessel head temperature on trend display. (Plant Computer points IITE2040 and IITE3040)3.8 IF lowering of RCS level is to continue, THEN DRAIN to between 47 ft. 4 in.and 46 ft. 5 in. elevation as indicated on RC-LG-3662 "RCS LEVEL SIGHTGLASS (SLINKY)".

This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 201 of 216 Plant Cooldown E Addendum 19 Controlling RCS Inventory at or above Elv 39 ft. 4.9 in. Page 7 of 10 INITIALS NOTE WHEN RVWL Sensor Point I has been uncovered, THEN indicated temperature will rise to about 750'F due to heating from the heated junction thermocouple.

3.9 VERIFY

water in reactor vessel head less than 180'F as indicated by Plant Computer display RC12 (8112).3.10 PLACE the Reactor Vessel head to pressurizer equalizing line in service as follows: 3.10.1 ENSURE the reactor vessel head venting manifold is connected per OPOP02-RC-0003, Addendum 1, Filling and Venting the RCS.{RCB on RV Head)3.10.2 ENSURE the RV to PZR Equalizing Line is aligned lAW OPOP03-ZG-0009, Mid-Loop Operation.

This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 202 of 216 Plant Cooldown Addendum 19 Controlling RCS Inventory at or above Ely 39 ft. 4.9 in. Page 8 of 10 INITIALS 3.11 WHEN RCS level is between 47 ft. 4 in. and 46 ft. 5 in. as indicated on RC-LG-3662 "RCS LEVEL SIGHTGLASS (SLINKY)", THEN PERFORM the following:

3.11.1 OPEN the following valves to remove water plug (Loop Seal) from Head and Pressurizer vent manifold and Rx Head Vent line: "1(2)-RC-0509 RX VESSEL HEAD""VENTING MANIFOLD DRAIN"{RCB On RV Head}"1(2)-RC-0507 RX VESSEL HEAD-"VENTING MANIFOLD VENT VALVE"{RCB On RV Head}3.11.2 IF the reactor vessel head vent valves are operable, THEN OPEN the Reactor Vessel head vent valves. {CP005}* "ISOL HV-3657A"* "ISOL HV-365713"* "ISOL HV-3658A" 0 "ISOL HV-3658B"* "HEAD VENT THROT VLV HCV-0601"* "HEAD VENT THROT VLV HCV-0602" 3.11.3 IF desired, THEN OPEN PRT "1(2)-RC-0025 N2 SUPPLY ISOL".{RCB 6 ft E of PRT}3.11.4 IF RCS level will be maintained, THEN REFER TO Step 3.22.This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 203 of 216 Plant Cooldown I Addendum 19 Controlling RCS Inventory at or above Elv 39 ft. 4.9 in. Page 9 of 10 INITIALS 3.12 IF RCS level is to remain below 47 ft. 4 in, THEN PERFORM the following:

CLOSE "1(2)-RC-0507 RX VESSEL HEAD""VENTING MANIFOLD VENT VALVE"{RCB On RV Head}ENSURE "1(2)-RC-0509 RX VESSEL HEAD" "VENTING MANIFOLD DRAIN"{RCB On RV Head} remains OPEN to vent the HEAD.NOTE The temperature rise will occur when sensor is uncovered prior to RVWL point indicating dry.WHEN RVWL Sensor Point I has been uncovered, THEN indicated temperature will rise to about 7507F due to heating from the heated junction thermocouple.

WHEN RC-LG-3662 "RCS LEVEL SIGHTGLASS (SLINKY)" is in service, THEN ENSURE "OPOP07-RC-0001, RC Vent Rig/Sightglass Installation and Removal", LINEUP 1. "RC-LG-3662 "RCS LEVEL SIGHTGLASS (SLINKY)" in Service Lineup" PERFORMED daily and DOCUMENTED in a temporary log.CAUTION Inadequate head venting due to rapid draining can cause Reactor Vessel water level to remain higher than loop level.3.13 IF lowering of RCS level is to continue, THEN DRAIN RCS until RVWL Sensor Point 1 (45' 3.4") heated junction thermocouple temperature rises by 207F. (Plant Computer Points IITE2004, A-Train, IITE3004, C-Train)3.14 COMPARE RCS level sightglass indication with RVWL level.3.15 IF RVWL AND RCS level sightglass do NOT agree within 6 inches, THEN STOP the drain and REFER TO Addendum 15, Rx Head Venting and RCS level Instruments Disagreements Evaluation Guideline.

3.16 IF RCS level will be maintained, THEN REFER TO Step 3.22.This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 204 of216 Plant Cooldown Addendum 19 Controlling RCS Inventory at or above Ely 39 ft. 4.9 in. Page 10 of 10 INITIALS 3.17 REFER TO Addendum 3, Determination of RCS Volume to be Drained, to determine draindown volume.NOTE* The temperature rise will occur when the sensor is uncovered prior to RVWL point indicating dry." WHEN RC-LG-3662 "RCS LEVEL SIGHTGLASS (SLINKY)" is in service, THEN ENSURE "OPOP07-RC-0001, RC Vent Rig/Sightglass Installation and Removal", LINEUP 1, "RC-LG-3662 "RCS LEVEL SIGHTGLASS (SLINKY)" in Service Lineup" PERFORMED daily and DOCUMENTED in a temporary log.* Do NOT Reduce RCS level Below Ely 39 ft. 4.9 in. using this Addendum.3.18 IF lowering of RCS level is to continue, THEN DRAIN RCS until RVWL Sensor Point 2 (39' 4.9") heated junction thermocouple temperature rises by 20'F. (Plant Computer Points IITE2007, A-Train, IITE3007, C-Train)3.19 COMPARE RCS level sightglass indication with RVWL level.3.20 IF RVWL AND RCS level sightglass do NOT agree within 6 inches, THEN STOP the drain and REFER TO Addendum 15, Rx Head Venting and RCS level Instruments Disagreements Evaluation Guideline.

3.21 ADJUST RHR HX flow as required to maintain RHR HX inlet temperature less than 140'F.3.22 MAINTAIN RCS level using any of the following:

• Raise RCS level as necessary using gravity drain from RWST through the LHSI pump cold leg injection valves in the idle RHR train* Raise RCS level as necessary using CCP normal charging or seal injection* Reduce RCS level as necessary using low pressure letdown 3.23 MAINTAIN temporary log OPOP07-RC-0001, LINEUP 1, "RC-LG-3662 RCS LEVEL SIGHTGLASS (SLINKY)" in Service Lineup as required. (Ref.Procedure Step 3.37)3.24 To raise RCS level, GO TO Step 2.0 of this addendum.3.25 RETURN TO Procedure Step in effect.This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 205 of216 Plant Cooldown Addendum 20 Venting Reactor Vessel Head Using Head Vent Throttle Page 1 of 2 Valve(s)INITIALS NOTE* This addendum vents a voided Reactor Vessel head when the RCS in Modes 5 or 6." HJTC Train "A" or Train "C" RVWL Sensor I indications are found at Computer Points IITE2004 and I1TE3004, respectively.

1.0 CHECK

the following indications for Reactor Vessel Head voiding:* Pressurizer level rising AND" VCT level constant OR rising AND* RVWL Sensor 1 temperature rising 2.0 IF Reactor Vessel Head voiding is indicated, THEN vent the Reactor Vessel Head to the PRT as follows: 2.1 OPEN Head Vent Isolation Valves:* ISOL HV-3657A and ISOL HV-3658A OR* ISOL HV 3657B and ISOL HV-3658B 2.2 OPEN Head Vent Throttle Valve(s):* HCV-0601* HCV-0602 This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 206 of 216 Plant Cooldown Addendum 20 Venting Reactor Vessel Head Using Head Vent Throttle Page 2 of 2 Valve(s)INITIALS 2.3 MONITOR for the following indications of the Reactor Vessel Head being vented: " Pressurizer level lowering AND* RVWL Sensor I temperature lowering AND* PRT pressure rising 2.4 WHEN the Reactor Vessel Head void is vented, THEN PERFORM the following:

2.4.1 ENSURE

CLOSED the Head Vent Throttle Valve:* HCV-0601* HCV-0602 2.4.2 ENSURE CLOSED the Head Vent Isolation Valves:* HV-3657A* HV-3657B* HV-3658A" HV-3658B This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 207 of 216 Plant Cooldown Addendum 21 Closure of Personnel Air Lock Doors Page 1 of 2 1.0 Purpose: 1.1 Provides instructions for closing Personnel Air Lock (PAL) doors to establish containment closure during work activities inside Reactor Containment Building (RCB) that require both PAL doors to be open.2.0 Instructions:

2.1 PERFORM

the following to close Personnel Air Lock doors (M-90): 2.1.1 ENSURE the following conditions at RCB PAL door:* Door-seating surfaces are clear of all obstructions.

• Handwheel position indication in "OPEN" position.* Both door latch pins fully retracted to ensure NO interference with door closure.2.1.2 PULL RCB PAL door CLOSED.2.1.3 WHEN RCB PAL door is flush with door jam, THEN ROTATE RCB PAL door hand wheel to "CLOSED" position to engage door latch pins.2.1.4 ENSURE the following conditions at MAB PAL door:* Door-seating surfaces are clear of all obstructions.

• Handwheel position indication in "OPEN" position.* Both door latch pins fully retracted to ensure NO interference with door closure.2.1.5 PULL MAB PAL door CLOSED.2.1.6 WHEN MAB PAL door is flush with door jam, THEN ROTATE MAB PAL door hand wheel to "CLOSED" position to engage door latch pins.This procedure, when completed, SHALL be retained.

I OPOP03-ZG-0007 Rev. 71 Page 208 of 216 Plant Cooldown Addendum 21 Closure of Personnel Air Lock Doors Page 2 of 2 NOTE An ECO Caution Tag will be hanging on 1 (2)-XC-0037 and MCB switch for PAL Seal OCIVs to prevent pressurizing door seals with the doors open.2.2 PERFORM the following to pressurize PAL door seals: 2.2.1 OPEN "1(2)-XC-0037 RCB PERSONNEL AIRLOCK SEAL AIR SUPPLY ISOLATION VALVE". (MAB 60', Room 326)2.2.2 NOTIFY Control Room (8614, 8610, 1111(7953, 8683, 2222)} that BOTH PAL doors are in the CLOSED position AND valve 1(2)-XC-0037 is OPEN.2.2.3 REQUEST Control Room to pressurize PAL door seals by taking"PERS AIR LOCK SEAL OCIVS" switch to OPEN. (CP005)This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 209 of216 Plant Cooldown Data Sheet 1 RCS and Pressurizer Cooldown Rates Page 1 of 4 Unit: Date: Initial Time: I. RECORD data at 15 minute intervals.

2. Pressurizer cooldown rate SHALL NOT exceed 200"F (160 0 F/hr ADMIN LIMIT) in any one hour. (TRM 3.4.9.2) Cooldown rate is the actual cooldown over the hour period.(i.e. :°FTemperature recorded 60 minutes prior to current time -°FTemperature current = A°F/hour)3. OBTAIN RCS pressure from QDPS Detail Data Menu Page 1 (PT403, PT404, PT405, PT406 or PT407)4. OBTAIN RCS temperature from QDPS Detail Menu Page 1 (TE414, TE424, TE434, or TE444)5. RCS cooldown rate SHALL NOT exceed 100°F (80'F/hr ADMIN LIMIT) in any one hour period during cooldown within the limits of Addendum 1. (Technical Specification 3.4.9.1) (Ref 2.70)Cooldown rate is the actual cooldown over the hour period. (i.e.: 0 FTemperatUre recorded 60 minutes prior to current time -0 FTeniperature current = A°F/hour)6. RECORD the differential temperature between the Pressurizer water space TI-0608 and RCS temperature (4). IF cooldown is due to an unisolable RCS leak, THEN differential temperature limit is less than or equal to 250'F. Differential temperature limit is less than or equal to 320'F for normal cooldowns. (Ref 2.62)7. INITIAL the appropriate column.8. 15 minute cooldown rate will be converted to an hourly rate.Calculation: (i.e.: (°FTemperature recorded 15 minutes prior to current time -°FTemperature current ) x 4 = A°F/hour)This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 210 of216 Data Sheet 1I RCS and Pressurizer Cooldown Rates Page 2 of 4 9. IF any readings obtained are outside the specified limits, THEN PERFORM the following: (Ref 2.7)9.1 Immediately NOTIFY the Shift Manager/Unit Supervisor of the out of specification reading.9.2 STOP the plant cooldown.9.3 ENSURE RCS temperature/pressure within the specified limits from the QDPS.9.4 MAINTAIN existing RCS temperature and pressure.NOTE The cooldown SHALL NOT be resumed until Engineering authorizes the cooldown to be restarted.

9.5 NOTIFY

Engineering to perform an Engineering Evaluation to determine the effects of the out-of-limit (NOT ADM1N LIMIT) condition on the structural integrity of the RCS or Pressurizer, as applicable.

10. WHEN Pressurizer vapor space temperature TI-0607 is NOT functional, THEN use the associated functional Pressurizer water space temperature TI-0608 for all Pressurizer temperature indications called out in this procedure.

Use of the liquid temperature element alone is more conservative

[will provide higher indicated change for a given actual system change] and better represents actual metal temperature.

Use of the liquid temperature indication alone will provide assurance that cooldown limits will NOT be exceeded. (CREE 02-3367)Example:* Pressurizer vapor space temperature TI-0607 is non-functional, THEN substitute Pressurizer water space temperature TI-0608, for Pressurizer vapor space temperature TI-0607 in this procedure.

This procedure, when completed, SHALL be retained.

IOPOP03-ZG-0007 Rev. 71 Page 211 of 216 Plant Cooldown I Data Sheet I RCS and Pressurizer Cooldown Rates Page 3 of 4 Unit:- Date: Initial Time: Page -of TIME PRZR VAPOR SPACE PRZR WATER SPACE RCS PRESS RCS PRZR-RCS WTR INITIAL (I) QDPS (3) (7)TI-0607 0 F/HR (8) 0 F/HR (2) TI-0608 0 F/HR (8) °F/HR (2) QDPS (4) oF/HR (8) °F/HR (5) Delta T (6)15 minute Rolling Hourly 15 minute Rolling Hourly 15 minute Rolling Hourly rate rate rate rate rate rate_ _ _ _ _ _......_._ a ._ _ _ _ .._ _ _ _Personnel Participating in cooldown: Name Initials Date Initials Name Cooldown completed by: Data Sheet 1 Reviewed by: Operator Time Shift Manager/Unit Supervisor Date This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 I Rev. 71 Page 212 of216 Plant Cooldown E Data Sheet 1 RCS and Pressurizer Cooldown Rates. Page 4 of 4 Unit: Date: Initial Tirne:_ Page__ of-(CONTINUATION SHEET)TIME PRZR VAPOR SPACE PRZR WATER SPACE RCS PRESS RCS PRZR-RCS WTR INITIAL ((10) QDPS (3) (7)TI-0607 °F/HR (8) °F/HR (2) TI-0608 °F/HR (8) °F/HR (2) QDPS (4) °F/HR (8) °F/HR (5) Delta T (6)15 minute Rolling Hourly 15 minute Rolling Hourly 15 minute Rolling Hourly rate rate rate rate rate rate This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 f Rev. 71 Page 213 of 216 Plant Cooldown I Lineup 1 RV to PZR Equalizing Line Lineup Page 1 of 2 UNIT 1 (Circle Unit Performing Lineup) UNIT 2 EXCEPTIONS DEVICE COMPONENT NOUN REMARKS NUMBER DESCRIPTION Personnel participating in device manipulation:

Name Initials Device lineup completed by: Operator Date Time Lineup I Reviewed: Unit Supervisor Date This procedure, when completed., SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 214 of216 Plant Cooldownj ineup 1 RV to PZR Equalizing Line Lineup Page 2 of'2 )DEVICE COMPONENT NOUN LOCATION POSITION ALIGNED VERIFIED NEW TAG NUMBER DESCRIPTION REQUIRED BY BY NEEDED RX VESSEL HEAD VENTING 1(2)-RC-0507 MANIFOLD VENT VALVE RCB On RV Head CLOSED 1(2)-RC-0509 RX VESSEL VENTING MANIFOLD RCB On RV Head ** OPEN/DRAIN CLOSED RV HEAD/PZR EQUALIZING LINE RCB 73' Outside On SG 1(2)-RC-0504 DRAIN VLV IA(2A) N Shield Wall CLOSED RX VESSEL HEAD ATMOSPHERIC 1(2)-RC-0132 VENT VALVE RCB On RV Head ** OPEN/CLOSED RX VESSEL HEAD VENTING 1(2)-RC-0508 MANIFOLD PZR EQUAL LINE ISOL RCB On RV head OPEN RV HEAD/PZR EQUALIZING LINE ISOL RCB 73' Outside On SG 1(2)-RC-0501 VLV IA(2A) N Shield Wall OPEN I(2)-RC-0163 PZR SPRAY LINE VENT VALVE RCB Top of PZR OPEN 1(2)-RC-0 103 PZR SPRAY LINE VENT VALVE RCB Top of PZR OPEN RX VESSEL HEAD VENTING

• OPEN/1(2)-RC-0506 MANIFOLD PI-3636 ]SOL RCB On Head CLOSED RCB 1(2)-RC-0070 RX VESSEL HEAD VENT ISOL RV LOCKED On RV Head inside OPEN shield wall* IF PI-3636 is installed, THEN OPEN "I(2)-RC-0506 RX VESSEL HEAD VENTING MANIFOLD P1-3636 ISOL".** OPEN IF RCS LEVEL below 47 ft. 4 in.This procedure, when completed, SHALL be retained.

I OPOP03-ZG-0007 Rev. 71 Page 215 of 216 Plant Cooldown Form 1 CVCS Line Boration Tracking Form Page 1 of 2 UNIT 1 (Circle Unit Performing Form)UNIT 2 INITIALS NOTE Any component and line NOT to be borated to 2800 ppm or greater requires Unit Operations Manager prior permission. (Form 1, Step 3.0)The volumes in Steps 2.0 and 3.0 of this form purge five dead leg volumes through the associated component and piping.With the RCS borated to greater than or equal to 2875 ppm, CVCS system design will mix the following lines ensuring the concentration will be 2800 ppm or greater prior to entering the RCS piping. (Reference 2.112)CVCS sections that do NOT Mixing volume require prior boration 1(2)-CV-FCV-0201, CCP IA(2A) recirc to VCT Volume Control Tank 1(2)-CV-FCV-0202, CCP 1B(2B) recirc to VCT Volume Control Tank 1(2)-CV-LCV-31 19, Aux Spray line Pressurizer 1(2)-CV-0255, CVCS charging discharge Charging pump discharge header to RCS FCV-0205 bypass including Regenerative Heat Exchanger 1(2)-CV-0206, CCP I B(2B) discharge bypass Charging pump discharge header to RCS including Regenerative Heat Exchanger 1.0 IF a component will NOT be borated, THEN PERMISSION from Unit Operations Manager is required, AND N/A components in Step 2.0 and 3.0 of this addendum NOT to be borated.2.0 ENSURE the following components borated to 2800 ppm or greater as follows: ENSURE CCP IA(2A) borated with 600 gallons of 2800 ppm or greater of borated water from VCT through CCP IA(2A) "DISCH ISOL MOV-8377A" per OPOP02-CV-0004.

ENSURE CCP IB(2B) borated with 600 gallons of 2800 ppm or greater of borated water from VCT through CCP I B(2B) "DISCH ISOL MOV-8377B" per OPOP02-CV-0004.

ENSURE 1(2)-CV-MOV-0003 and associated lines to RCS borated with 300 gallons of 2800 ppm or greater borated water per Step 7.12 of this procedure.

ENSURE 1(2)-CV-MOV-0006 and associated lines to RCS borated with 300 gallons of 2800 ppm or greater borated water per Step 7.12 of this procedure.

US/SM This procedure, when completed, SHALL be retained.

0POP03-ZG-0007 I Rev. 71 IPage216 of Q 216 Plant Cooldown Form I CVCS Line Boration Tracking Form Page 2 of 2 3.0 RECORD Time/Date when each component and associated lines were determined to be borated.CCP IA(2A) and associated lines have been inservice with 600 gallons of 2800 ppm or greater borated water flushed.CCP I B(2B) and associated lines have been inservice with 600 gallons of 2800 ppm or greater borated water flushed.1(2)-CV-MOV-0003 and associated lines to RCS have been inservice with 300 gallons of 2800 ppm or greater borated water flushed through it.1 (2)-CV-MOV-0006 and associated lines to RCS have been inservice with 300 gallons of 2800 ppm or greater borated water flushed through it.////This procedure, when completed, SHALL be retained.

RS1 CALC. NO.STPNOC013-CALC-002)E N E R C 0 N CALCULATION COVER SHEET REV. I ExrA)tence AEer-y wro.t Every dft PAGE NO. 1 of 49 Title: Radiological Release Thresholds for Emergency Client: South Texas Project Action Levels Project: STPNOC013 Item Cover Sheet Items Yes No 1 Does this calculation contain any open assumptions that require confinration? (If YES, Identify the assumptions)

" 2 Does this calculation serve as an "Alternate Calculation"? (If YES, Identify the design verified Calculation.)

Design Verified Calculation No. /3 Does this calculation Supersede an existing Calculation? (If YES, identify the superseded Calculation.)

Superseded Calculation No. V Scope of Revision:

Incorporate decay time of one hour from shutdown as well as migration into Attachment

1. Change statement of no decay in the STAMPEDE runs.Revision Impact on Results: Values calculated in Attachment I decreased.

and have become the limiting values.J m"" m F ~tudr-alcaflati n-I--- R=imrl rleu irtIn=z F L1I Safety-Related E Non-Safety Related D (Print Name and Sign)Originator:

Caleb Trainor Date: 3/21/2014 Design Verifier:

Chad Cramer Date: 3/21/14 Approver:

Marvin Morris Date: 3/21/14 CALC. NO. STPNOC013-CALC-002 ENERCON C CU TONREV. 1 REVISION STATUS -SHEET R, PAGE NO. 2 of 49 CALCULATION REVISION STATUS, REVISION DATE DESCRIPTION 0 03/0M3/2014 Initial Issue 1 3/21/2014 Resolve inconsistency in decay times for the two calculations PA'.GE REV1ISiON STATUS'PAGE NO. REVISION .REVISION AT'TACHMENT REVISION STATUS ATTACHMENT NO.PAGE NO. REVISION NO. ATTACiMENT NO,:. PAGE NO.REVISION NO.2 1" 2 12-24., 25-31 32-49.1 3.C, CALCULATION ON CALC. NO. STPNOC013-CALC-002 C ENE RCON DESIGN VERIFICATION REV. ICKLIST ..PAGE NO. 3 of 49 Item CHECKLIST ITEMS Yes No N/A 1 Design Inputs -Were the design inputs correctly selected, referenced (latest revision), consistent with the design basis and incorporated in the calculation?

2 Assumptions

-Were the assumptions reasonable and adequately described, justified and/or verified, and documented?

3 Quality Assurance

-Were the appropriate QA classification and requirements V/assigned to the calculation?

4 Codes, Standard and Regulatory Requirements

-Were the applicable codes, standards and regulatory requirements, including issue and addenda, properly V" identified and their requirements satisfied?

5 Construction and Operating Experience-Have applicable construction and V/operating experience been considered?

6 Interfaces

-Have the design interface requirements been satisfied, including v/interactions with other calculations?

7 Methods -Was the calculation methodology appropriate and properly applied to_-satisfy-thetcalculation-objective?--

......8 Design Outputs -Was the conclusion of the calculation clearly stated, did it correspond directly with the objectives and are the results reasonable compared to V/the inputs?9 Radiation Exposure -Has the calculation properly considered radiation exposure to the public and plant personnel?

10 Acceptance Criteria -Are the acceptance criteria incorporated in the calculation sufficient to allow verification that the design requirements have been V/satisfactorily accomplished?

11 Computer Software -Is a computer program or software used, and if so, are the requirements of CSP 3.02 met?COMMENTS: None-(PrintJVame and Sign)Design Verifier:

Chad Cramer Date: 3/21/14 Others: Date:

ENERCON Exce~femte-EwilypMfort Everyft CALCULATION.

DESIGN VERIFICATION PLAN AND

SUMMARY

SHEET CALC. NO. STPNOC013-CALC-002 REV. 1 PAGE NO.4 of 49 Calculation Design Verification Plan: Calculation shall be verified by comparing the documented input with the references and checking the validity of the references for the intended use. As necessary, assumptio ns shall be evaluated ind verified to determine if they are based on sound engineering principles and practices.

Verify the applicable metoldology, inputs, results, and cbnclusions..(I (Print Name and Sign for Approval -mark "NIA if nhot reijUlred)

Appiove: Date: 8)21/1 Approver:

Marvin Mor'ris :.: Calculation Design Verification Siummary":

... ....._.._ _- ._Design inputs, assumptions, methodology, results and conclusions .were evaluated/verified and found to be acceptable.

All comments have been incorporated.

Based On The Above Summary, The Calculation Is Determined To Be Acceptable.

Ci (Print Name and Sign)Design Verifier:

Chad Cramer Date: 3/21/14 Others: Date: C.

CALC. NO. STPNOC013-CALC.002 E 4 E R C 0 N Radiological Release Thresholds REV.I for Emergency Action Levels ExceiM,#-Eeiy ro veit. Ety dAy PAGE NO. 5 of 49 Table of Contents 1.0 OBJECTIVE/SCOPE

...........................................................................................................

6 2.0

SUMMARY

OF RESULTS ..................................................................................................................

6 3.0 METHOD OF ANALYSIS ..........................................................................

.................................

7 4.0 IN P U T S ................................................................................................................................................

7

5.0 REFERENCES

......................................................................................................................................

7 6.0 ASSUMPTIONS

..............................................................

I ....... 8 7.0 STAMPEDE CALCULATIONS

...................................

6 .................................................................

9 7.1 U nusual E vent -R U I ...................................................................................................................................

9 7.2 Alert, Site Area and General Emergencies

-RA 1, RS1, RGJ ...........................

10 Attachment 1 -Hand Calculations

..................................................................................................................

12 Attachment 2 -Calculations

..............................

....................................

...............................................

25 Attachment 3 -STAMPEDE OUTPUT ...........................................................................................................

32 CALC. NO. STPNOC013-CALC-002 0 E NE CO N Rdio ogil Releas oThesholds 6of4for Emergency Action Levels~PAGE NO. 6 of 49 1.0 OBJECTIVE/SCOPE The purpose of this calculation is to determine the Emergency Action Level (EAL) threshold values of a radiological release from the Unit Vent or Main Steam Lines for an Unusual Event, Alert, Site Areý'Bmergency,'

or General Emergenicy.

The calculated threshold values ae to be included in the STP EAL Technical Basis document, which implements the new NEI 99-01, Revision 6, Emergency Action Level Schetne and will be submitted to the NRC for approvýal.

pon NRC approval, the values will be Used in 0ERPO 1-ZV-INO 1, Revision 10, Emergency Classification.

Both a hand calculation and the South Texas Assessment Model Projecting Emergency Dose Evaluation (STAMPEDE)

Softwre program weere 'ised to generate the results. The hand calculation is included as Attachment 1.Revision I of this calculation incorporated decay for a release taking place one hour after reactor shutdown.

This 'was done to create continuity between the two methodologies pesent.2.0

SUMMARY

OF RESULTS'The results of the calJulations for the radiation monitors specified in the STP EAL Basis Document and are listed in Table 2.1, below.Table 2.1: Summary of Calculation Results C Emergency Action Level RT-8010D, Unit Vent (pCilsec)RT-8046 th Main: Ste (PiC RU1 I Unusual Event I Alert.i rough 8049, am Lines f i/cc)tht- MC IýVxz Y I*STAMPEDE was not used to determine the threshold for RUI methodology should be used to determine the threshold value.This calculation will be used to establish the threshold values for abnormal radiation based emergencies in the STP EAL Technical Basis document.

.

ALC. NO. STPNOC013-CALC-002 0 E N E R C O N Radiological Release Thresholds REV. I for Emergency Action LevelsN b~ne-w PM W~y ftPGEN.7 f4 3.0 METHOD OF ANALYSIS Previously, STAMPEDE was used to calculate the Emergency Action Level threshold values for effluent releases.

A hand calculation will verify the STAMPEDE calculations.

The hand calculation is described in Attachment 1 of this document STAMPEDE conforms to the requirements of STP Procedure OPGP07-ZA-0014, Software Quality Assurance Program. STAMPEDE was run at STP on an STP computer and under the supervision of an ENERCON employee with access to the STP site as a critical worker.4.0 INPUTS 4.1 Per NEI 99-01, Revision 6, Initiating condition AU1, EAL 1, the Notice of Unusual Event initiating condition is a release of gaseous or liquid radioactivity greater than two times the ODCM limit for sixty minutes or longer (Reference 5.10).4.2 The ODCM .offsite dose limit is exceeded if the Xe-133 release concentration exceeds 7.41E-04 piCi/cc (Reference 5.6).4.3 The Unit Vent flow rate is 9.4E+07 cc/sec (Reference 5.1).4.4 The main steam line pressure and PORV choke flow rate are 1285 psig and I .05E+06 lbmlhr, respectively (Reference 5.2).-4-.5---The-spei-fic-volume-ofosaturated-stear-ar--V-285-psig i,0;3-38-ft -tlb-m=(R~

f 4.6 The concentration is varied to find the release concentration which correlates to each emergency action level. Emergency action levels are taken from NEI 99-01, Revision 6 (Reference 5.10) for initiating conditions AA1, AS 1 and AGI. EAL I is the EAL of interest in each initiating condition, The doses at the Site Boundary that correlate to the threshold concentrations are listed in Table 4.1.Table 4.1 EAL Offsite Dose Initiating Conditions Alert Site Area General

5.0 REFERENCES

5.2 Main Steam PORV Capacity Verification MC05591, Revision:

1 5.3 NIST Steam Tables, 2011 5.4 OERP01-ZV-INO1, Emergency Classification Draft Revision 10 5.5 OERPO 1 -ZV-TPO 1, Offsite Dose Calculations, Revision 21 5.6 STP Calculation NC-9012, CRMS Rad Monitor Setpoints, Revision 7 537 STP Calculation NC-901 1, Revision 2 5.8 STAMPEDE Computer Program, Revision 7.0.3.3 5.9 STAMPEDE User's Manual 5.10 NEI 99-01, Revision 6, Development of Emergency Action Levels for Non-Passive Reactors 5.11 OPGP07-ZA-0014 Quality Assurance Program :j 5.12 ITWMS Call Number 1000010987 Design Document, Revision 0 CALC. NO. STPNOCOI3-CALC-002 0 "E N E R C.O N Radiological Release Thresholds REV...: .. -for Emergency Action Levels r*Yf*t PAGE NO. 8 of 49 6.0 ASSUMPTIONS 6.1 Unit Vent Noble Gas Monitor To-be consistent'with the ODCM methodology, the.unit vent release is assumed t6be entirely Xe-133. The unit vent noble gas. monitor is calibrated to Xe-133 (Refere,*nc'5, 1) therefore; the monitor reading accurately reflects the Xe-133"release magnitiude.

To be consistent with ODCM methodology, the main steam line release is assumed to be entirely Xe-133. The noblegas monitor is calibrated to Xe- 13 3 (Reference 5.6).6.2 Release Duration Per Reference 5.10, S ections IC AAI.,. AS 1,and AGi developer notes, the release should be assumed to last on6 hour..6.3 Release folilowingReactor Shutdown The release initiates one hour after reactor shutdown.

While a release initiating at'reactor" shutdown is likely, signifiddrit decay of short lived nuclideg occurs during the mligration time. A release at reactor shuiitdown would have a significantly higher activity at the monitor location than

'site.

this- -.--. ---would a very high threshbid Which .would not bb-appropriate for releases which occur shortly after shtiidown.

One hour adr reactor shtýtaow is sufficfent time to decay short lived nuclides and createoa conservatie threshld..

6.4 Source

Term Per Reference 5'1, any' unit vent releaset itlfh increed RCS act ivity and no core melt should be calculated using tlhe-gap inventory.

Thedftore the gap in,ýentoiy is used for all.unit vent releases.Per Reference 5.1 for a main steam line release followirg a steam generator tube rupture it is appropriate to use an inventory of noble gases plus 0.2% iodine. A steam generator tube rupture is the only scenario which would create significant offsite doses through a main steam line release.6.5 Default STAMPEDE Input Values Keference D. I d pr tes rtiatig c P *A-l-.A-St-and-A-&l-suge st-using-he ODCM or the site's emergency dose assessment methodology.

STAMPEDE is Wsed for emergency dose assessment.

Per Reference 5.1, when actual meteorology is not available, the default STAMPEDE values should be used. Had the ODCM methiaddolog6y been* used, the 500 hour0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br /> peak X/Q value would be used whichis less conservative than-the Xi/Q value produced by STAMPEDE using default meteorological c6ndition:

Therefore, the use of STAMPEDE default values provides a more conservative estimate than that of the altnriiative method outlined in Reference 5.10.6.6 Average Effluent Concentration (X/Q)The same X /Q is used for the unit vent and main steam line release. Reference

5.1 applies

the same unit vent X/Q to Units 1 and 2 which would also be applicable to the main steam line. All releases are considered to be ground level releases.

CAI.O. NO. STPNOCO1 3-CALC-002 EINNE R C 0 N Radiological Release Thresholds REV. I for Emergency Action Levels R'.ftolance-Evetvpfo Evamdox PAGE NO. 9 of 49 7.0 STAMPEDE CALCULATIONS

7.1 Unusual

Event -RU1 7.1.1 Unit Vent Monitor AUI recommends declaring an unusual event due to a release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or longer (Reference 5.10).STP sets the ODCM limit at 7.41E-04 j, Ci/cc (Reference 5.6, pg. 16). Two times the limit would be 1.48E-03 ýtCi/cc..

The threshold is listed in .iCi/sec so that variations in flow rate do not change the threshold.

The normal flow rate from the unit vent is 9.4E+07 cc/sec (Reference 5.1)., cc ~ ~Ci Concentration

• Flow Rate = Release Rate ccsec \ see/-(1.48E -03) (1 *,(9.4E + 07)(c)= 1.4E + 05 (!C§.. Equiatid-7.1.1:1.........71 7.1.2 Main Steam Line Monitor The ODCM does not calculate a release corresponding to allowable limits for the main steam line monitors.

Since the unit vent release calculated in the ODCM was assumed to be primarily Xe-133, the assumption is made in the ODCM that other noble gases and iodine may be ignored in the calculation.

This assumption is equally justifiable for the main steam line and the same limiting release will be used.The magnitude of the release calculated for the unit vent Unusual Event applies to the main steam lines as well. The main steam line PORV's will create a dose exceeding two times the ODCM limit by releasing 1.4E+05 p.Ci/sec of activity which is equivalent to the release from the unit vent.The steam lines hold saturated steam at 1285 psig, per Reference 5.2, which has a specific volume of 0.338 fl 3/lbm (Reference 5.3). The PORVs will release the steam at 1.05E+06 lbm/hr per Reference 5.2. This creates a set flow rate of steam from the main steam lines of 2.79E+06 cc/sec as shown below.flbm (ft 3\ CC ( sec~ cc F D Density.

• 28316.846

+ 3600* *= 279 + 6)lbm (t 3 (cc' (sec\ cc 1.OSE +06 hr*) 0.338 28316.846

+ 3600 -)= 2.79E + 06-Ehr aionbm 7.. sec Equation 7.1.2.1 O " " .CALC. NO. STPNOC013-CALC-002 Radiologic.al Release Thresholds CL. N.SMO03CL-

'E-NER.CO.N REV. I.d .. -for Emergency Action Levels PAGENO. 10 of 49 Since the flow rate is set, the concentration will determine the limit. Equation 7.1.1.1 solves for the limiting concentration of 5.OOE-02 1 , Ci/cc as shown below.Limiting Release (Jr!) _ ., sec = Limiting Concentration (Release Rate ccC/-* .oo 06 02 ( : 2.79* 106 cc (\cc Equation 7.1.2.2 7.2 Alert. Site Area andGeneral Emergencies

-"RAl, RS1, RGI.7.2.1 Unit Vent Monitor C Input The Alelr EAL is set to 10 mrem TEDE and 50 mrem Thyroid CDE per Reference 5.10._- The-emergency-offsite-dose.

cdlculation-softwae STA -EDE. was _wdJQ kuja~te the-release which corresponds to this dose. A release concentration correlating to the EAL threshold value was calculated by varying the input. The foll6wing assu.mptions and inputs were -used forthe calculation as described in Sections 4.0 and 6.0..Release begins at reactor triP..* Release lasti for qie hour:.G Gap surce te" .* Defauklt STAMPEDEVinput.values.

o :Windspeed=

13.2 mph o StabilityW class D.Results'C Given a monitored unit vent release of 2.50E-i+06 PICi/sec, the.Thyroid CDE is 51 andt-he-EATInitiatingCondition i.exceeded., Threshold values for the Site A.rea Emergency ay d General Emergenclis are multiples of 1 0 and 100 of the AIeit. Sincethe correlatioinbetween release cojncentration and dose is linear, threshold values for the steam line monitors are 2.50E+07 and 2.50E+08 [LCi/sec for the SAE and GE:respectively.

Both. are also limited by Thyroid CDE. Additional STAMPEDE iterationis Were performed to confirm this and aie attached.The input and output files can be found at the end of this document in Attachment 3.C CALC. NO. STPNOCOI3-CALC-002 EN E R CO N Radiological Release Thresholds for Emergency Action Levels REN. IAW, ydaX PAGE NO. I I of 49 7.2.2 Main Steam Line Monitor Input A release concentration correlating to the EAL threshold value was calculated by varying the input. The following assumptions and inputs were used for this calculation as described in Sections 4.0 and 6.0." Release begins at reactor trip* Release lasts for one hour* Noble gas + iodine with 0.2% iodine source term* Default STAMPEDE input values o Windspeed

= 13.2 mph o Stability class D Results Given a monitored main steam line release of 4.5 ýiCi/cc, the Thyroid CDE is 50 mrem/hr and the EAL Initiating Condition is exceeded..-..... --Tie-inputtand-output-files-can be found-at-the end of-this documentin Attachment3;

.- -.... -7.3 Threshold values for the Site Area Emergency and General Emergencies are multiples of 10 and 100 of the Alert. Since the correlation between release concentration and dose is linear, threshold values for the steam line monitors are 45 and 450 4tCilcc for the SAE and GE respectively.

Both are also limited by Thyroid CDE, Additional STAMPEDE iterations were performed to confirm this and are attached.

Radiological Release Thresholds ALC. NO. STPNOC013-CALC-002 0 E N E R C O N for Emergency Action Levels REV._Attachment 1 PAGE NO. 12 of 49 I Attachment 1 -Hand Calculations 1.0 OBJECTIVE/SCOPE Each release calculated using STAMPEDE in the main document is calculated by hand in this attachment and the results compared to STAMPEDE.2.0

SUMMARY

OF RESULTS Table 2.1 is displayed again below showing the'results from all the calcuIations.

The minor difference is due to STAMPEDE using decay factors over a one hour period after shutdown.

This also accounts for the change in the limiting dose being TEDE in the hand calculations and Thyroid CDE in the STAMPEDE calculations.

The accuracy of the hand calculation is considered sufficient and recommended for use in Emergency Action Levels.Table'2.1 Results Emergency Action RT-8010b, Unit Vent_- ___- _ .. .

RT-8046 through 8049,.Main Steam-Line-

--* (plci/cc)C HAI 1151 mI 6e Area E~mergencyI I uenerai Eimergency 1 3.0 METHOD OF ANALYSIS Using the limiting dose at the site boundary, the release is back calculated using atmospheric dispersion models. The X/Q value used is calculated from Regulatory Guide 1.145,.Atmospheric Dispersion Models for PotentialAccident Consequence Assessments at Nuclear Power Plants. Rather than using the most conservative meteorology, average meteorological conditions are used as inputs C Radiological Release Thresholds CALC. NO. STPNOC013-CALC-002 O NE RCO N "for Emergency Action Levels RV. 1 Attachment 1 PAGE NO. 13 of 49 to most closely agree with STP emergency dose assessment methodology per the ODCM and STAMPEDE.

Assumed nuclide inventories are taken from Reference 5.4. The dose conversion factors are taken from Reference 5.2. A release concentration is used to find an initial projected dose at the Site Boundary.

Using the projected dose at the Site Boundary, the release concentration is scaled to find the limiting dose for each EAL.4.0 INPUTS* The Unit Vent flow rate is taken from the Offsite Dose Calculation Manual; Revision 17, March 2011 and is 9.44E+07 cc/sec.* The main steam line pressure and PORV choke flow rate were taken from Reference 5.5 and are 1285 psig and 1.05E+06 lbm/hr respectively.

• The specific volume of saturated steam at this pressure is taken from the NIST steam tables and is 0.338 ft 3/Ibm.* The release concentration is varied to find the release concentration which correlates to each emergency action level dose. Emergency action level doses are taken from NEI 99-01 Revision 6 for initiating conditions AA 1, AS 1 and AG1. EAL 1 is the EAL of interest in each initiating condition.

The limiting doses are listed in Table 4.1. NEI 990-01 Revision 6 states that these values -bWA-tio--.

---

Guidelines (EPA PAGs) and the General Emergency represents the protective action values recommended by the EPA.Table 4.1 EAL Thresholds Alert Site Area General* A release lasting one hour is selected per NEI 99-01 Revision 6 developer notes.* Atmospheric dispersion factors are calculated per Regulatory Guide 1.145 (Reference 5.1). The reactor building dimensions used as inputs for this calculation are taken from Reference 513.* Nuclide inventories are taken from TGX/THX 3-1, (Reference 5.A) which is the source document for the nuclide inventories used in STAMPEDE.

The release inventories are a gap release and noble gases plus 0.2% iodine which are listed below. Each nuclide inventory was normalized to 1r,4ZV dULIVLUU5.

Radiological Release Thresholds CALC. NO. STPNOC013-CALC-002 I0 E N ER C O0 N for Emergency Action Levels IE .4 Attachment I PAGE NO. 14 of 49 Table 4.2 Gap InventoryActivity Normalized Nuclide' Actiifty (.wiClk Normalized 1-132 l.50E+/-+05 1.53E-03 Xe-137 ~1.00 +O7 1.747,75OI 1-134 2.4:0t+05 2.45E-03 Cs-'134 3,7'0E;ý+ý01 3.78E07 Kr-83mn 1.36F-+06 l.33E-ý021 Tel32., 4.0Ei 4 91E-108 Kr-85 370E+/-95 3.78E0 Ru ........3 9.00E-l11 IXr-88 7.g6tEý66

719EO PS9 1.10E-02' 1.12EM,'Xe-131 -l1OB.5 .ll~Q .T Ce44 A.O-3 75E Xe-133 2.tE+7 2ý25P,-Ol Sr9 AG-0 55-lo A,{-C C Table 4.3 Noble Gases+0.2%

Iodine Inventory Nuclide Inventory,-Nhrmlied 1-,13. .6E0 92P19E0 1-13 I861-oý2 6.9E05'"x in3 2.80E+OO 1jA.04E0 Xe-13n~ 770E+O 7 1776E02:.d~X.~J~i 1.JV.LU I .'t.L IU X6-138 5.80E'01 2.15E-03 Kj.r-85 7.60E+OG -2.82E-02 Kr-97 9.80E-01 3.63E-03 Kr-89 8.40E-02 3.12E-04 The dose conversion factors taken from EPA 400R92001 (Reference 5.2) are listed in Tables 4.4 and 4.5 below.C Radiological Release Thresholds CALC. NO. STPNOCO13-CALC-002 0 E NE R CON for Emergency Action Levels REV. I& ,,n -Attachment 1 PAGE NO. 15 of 49 Table 4.4 TEDE Dose Conversion Factors Dose Conversion Factor Nuclide (rem Der uCi*hr/cc)

Dose Conversion Factor (rem per uCi*hr/cc)

Nuclide 1-1324.90E+03 Xe-137 lE0 1-134 3. 1OE-I03 Cs-134 6.30E+04 Kr-85 1.30E+00 Ru1O3 13E0-88 1.30E+03 Zr95 32E0 Xe-l3tin--- -4.9----Ce144 -----4M.0E+05 3 I 2.OOE+O Sr89 5.00E+04 Table 4.5 Thyroid CDE Dose Conversion Factors Thyroid CDE DCF Nuclide (rem Cer uCl*hr/cc)

The unit vent noble gas monitor energy efficiency by nuclide is taken from Offsite Dose Calculation Manual (Reference 5.3). The values are relative to Xe-133 efficiency since the monitor is calibrated to Xe-133. Table 4.6 displays the energy efficiency by nuclide relative to Xe-133.

Radiological Release Thresholds CALC. NO. STPNOCOI 3-CALC.002 0 E N E R C O N for Emergency Action Levels REV. I Attachment 1 PAGE NO. 16 of 49 Table 4.6 Energy Efficiency Relative to Xe-133 Efficiency Relative to X9-133., e133m 0.1* " xv- ... '."0042 *There ish, .o relative efficiency available for Ass umption 6A4 fiurther justifies th omission.Table 4.7 Nuclide Half Lives Nuclide Half Life Nuclide Half Life (b~r)(hr)C-T-ý3 Mý =639PI-02 1-132 8Eý001 Xe-137 1-134 8.77E-01 Cs-134.1.80E÷04 1.83E+OO Te132 7.79E+01 Kr-85 9.40E+04 Ru103 9.44E+02 Kr-88 2.84E0O Zr-95 1,55E+03 Xe-131m 2.83E+02 Ce144 6.92E+03 Xe-133 1.27E+02 Sr89 1.21E+03 ci* The half-lives are taken from Reference 5.15 which lists the input data used by STAMPEDE.

Radiological Release Thresholds CALC. NO. STPNOC013-CALC-002 E N E R C 0 N for Emergency Action Levels REV. I FX6flc.c&U>'prqjCd ,iwydax Attachment 1 PAGE NO. 17 of 49

5.0 REFERENCES

5.1 Regulatory

Guide 1.145, Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants, Revision 1, November 1982.5.2 EPA 400R92001, Manual of Protective Action Guides and Protective actions for Nuclear Incidents, Revision 1, May 1992.5.3 Offsite Dose Calculation Manual, Revision 17, March 2011.5.4 TGX/THX 3-1, Revision 5, Westinghouse Radiation Analysis Manual.5.5 MC05591, Main Steam PORV Capacity Verification, Revision 1.5.6 NIST Steam Tables, 2011.5.7 OERPO1-ZV-INO1, Emergency Classification, Revision 10.5.8 OEIRP01-ZV-TPO1, Offsite Dose Calculations, Revision 21.5.9 STP Calculation NC-9012, Process and Effluent Radiation Monitor Set Points, Revision 7 5.10 STP Calculation NC-9011, CRMS Rad Monitor Setpoints, Revision 2.5.11 STAMPEDE Computer Program, Revision 7.0.3.3.5.12 STAMPEDE User's Manual 5.13 STP Drawing 6C1 89N5007, General Arrangement Reactor Containment Building, Revision 6 5.14 NEI 99-01, Revision 6, Development of Emergency Action Levels for Non-Passive Reactors.5.15- -IT'WMS Call.Numbor_1,!000-!0987-Design-Docjumenty-RevisionU 0 -.--- .... -6.0 ASSUMPTIONS

6.1 Release

lasts for one hour Per NEI 99-01 (Reference

.5.14), IC AA1, ASI, AG1 developer notes, the release should be assumed to last one hour.For this to be true for the main steam line, it is assumed that the PORV is open for one hour. To calculate the most limiting case, it is assumed that the maximum flow possible is being released from the PORV.6.2 Nuclide mix Per OERPO1-ZV-TPO1, Offsite Dose Calculations (Reference 5.8) any unit vent release with increased RCS activity and no core melt should be calculated using a gap inventory.

It is conservative to assume an increased RCS activity and not within the intended scope of the relevant initiating conditions to assume core melt. Therefore, a gap inventory is used for all unit vent releases.Per OERPOI-ZV-TP01, Offsite Dose Calculations (Reference 5.8) for a main steam line release following a steam generator tube rupture it is appropriate to use an inventory of 100 percent noble gases plus 0.2 percent iodine. Since a steam generator tube rupture releasing through the PORVs is the only steam generator tube rupture scenario which would create offsite doses large enough to meet or exceed the EALs, this assumption is made.

Radiological Release Thresholds CALC. NO. STPNOCO13-CALC002 E N E. R C O :N for Emergency Action Levels REV.1* Attachment 1 PAGE NO. 18 of 49 6.3 Atmospheric Dispersion NEI 99.01 (Reference 5.14) developer notes for initiating conditions AA 1, AS I and AGl suggest using the ODCM or the site's emergency dose assessnient methodology.

1-ZV-TP01, offsite Dose Calctilations (Reference 5.8), whien actoal-.neteorologyisnot available, the default.STAMPEDE values should be used. The default STAMPEDE valuesassume s stability class D for atmospheric dispersion and a windtpeed of 1.3.2 mph. These"Value$

were used as inputs for the atmosphe:ic dispersionrcaliulation..-...:

It is clear that STAMPEDE uses the same method for calculating atmospheric dispersion factor (X/Q) outlined in section 7.1. 1. of this Attachjment.

However, STAMPEDE, does not follow the same logic in selecting the .e."rstilt-from..

e three calculations.

The.ST EDE value printed in the"results found in attachm fit 3 is',consistent with the lag.get of the three hand calculated X/Q values. This gsts that STAMPEDE simiply seleets ihe laigest of the three X/Q values resulting in a much more consefVativ estimate, This calculatio.n wii deyiate from the recommendations of Regulatoi:yGuid.

1.145 and.donfdrm to the methodology STAMPEDE uses.The. close'proximity of all re dde points allows ftora single atmospheric dispersion coefficient to be used. This assustmption is also , .... ...-: (6.4 Exposure Pathways The dose conversion factors used in table 4.4 and 4.5 represent a summation of d'se conversion factors for external plume exposure, inhalation from theplume, and external exposure from deposition, Because the dose estimarlatos are used for implementing early phaseprotective actions, conv ersion .factors using limited pathliwys:

appropriate

..The EPA does not provide a dose conversion factor for kr-83m. Becausethe PAGs are based on-. EPA dos. calculations, it is appropriate to only us.-e th e nuclideg for which dose conversion factors areprvidbed.:-

Additionally, Kr-83m. represents only 1.33 -o of tie niclide inventory activity and its exclusion would not significantly affect the final dose.6.5 The release initiates one hour after reactor shutdown.

While a release initiating at reactor shutdown is likely, significant decay of short lived nuclides occurs during the migration time. A-!r~ o Oi gl

.ruterr-1 o *ation-thm-i-a at the re:eption Site. It isimpotit fotthethreshold to not be calculated at shutdown as this wo0uld create a very high t h res h old which w oul.d not be api ropriate for releages which occur sotyaer shutdown.i on'ehýur:

after reactor shutdown is shfiý n ie.od~i hr ie shortly after std n.OehuaferecosudonIs sufficie'nt tim~e to doe: 4y short lived nudides and createa conservativethresliold..." Decay is incorporated forone hour from reactor shutdown as well as migration time. Half-lives are taken:from Reference 5.15. Migration time is assuimed to be the reciprocal of the wind speed..

Radiological Release Thresholds CALC. NO. STPNOCO13-CALC-002 3 E E E R C 0 N for Emergency Action Levels REV. I imd Attachment I PAGE NO. 19 of 49 7.0 HAND CALCULATIONS 7.1 Unit Vent Monitor 7.1.1 X/Q The atmospheric dispersion factor, X/Q, determines the change in concentration between the unit vent discharge and the dose reception site. This value is based on meteorological conditions and will vary with wind speed and stability class. The ODCM uses the highest annual average X/Q value at the site boundary which is 5.3E-06 sec/m 3.However, for an accident related release STAMPEDE is used rather than the ODCM. STAMPEDE uses real time, user entered, or default meteorological conditions to calculate the X/Q for a specific accident.

Default values will be used as inputs into the Regulatory Guide 1.145 method for calculating X/Q as described below. Default values are identified in section 6.0, Atmospheric Dispersion.

For a neutral atmospheric stability class, which is the default in STAMPEDE, X/Q values can be determined through the following set of equations.

Q U 1 0 (ryr 2 Equation 7.1.1.1 X 1 QUio(3iwyo.)

Equation 7.1.1.2 X 1 Equation 7.1.1.3 Where X/Q = relative concentration (sec/m^3)it = 3.14159 U 1 0 = windspeed at 10 meters above plant grade (m/s)y = a er p me prea imn, a unc ion of atmosphTinc stabi ity and distance, determined from Regulatory Guide 1.145 Figure 1 Oz = vertical plume spread (in), a function of atmospheric stability and distance, determined from Regulatory Guide 1.145 Figure 2 y =(M -1)uy8oom + ouy = lateral plume spread with meander and building wake effects (m), a function of atmospheric stability, windspeed U 1 0 , and distance; M is determined from Regulatory Guide 1.145 Figure 3 A = the smallest vertical-plane cross-sectional area of the reactor building (mA2), taken from Reference 5.13 and shown below Radiological Release Thresholds tALC. NO. STPNOC013-CALC-002 E N EW R C O :N for Emergency Action Levels .REV. I Attachment 1 PAGE NO. 20 of 49 Figure 7.1.1.1: Reactor Building Dimensions C EL 241*'0 C Assuming the reactor building cross section to be a perfect rectangle and half sphere, the variables are defined as follows;U 1 0= 13.2 mph 5.9 m/s-z =4.2 m Fy (M -1)oy m + cy ;M=I -+ y =1200m A (135'* 158') + (L72) =-31128.37 The three equations become;x 1 2 5.9 7r120 *4.2

• 31128.37 C Radiological Release Thresholds CALC. NO. STPNOC013-CALC-002 ENERCO N for Emergency Action Levels R 1xle ,,-vv Attachment 1 PAGE NO. 21 of 49 X 1= 3.568

• 10-6 Q 5.9(37r

• 1200

• 4.2)x = 1= 1.07

• 10-s Q 5.9

• ir * [(1 -1)ayaoom + 1200]

• 4,2 To select the appropriate X/Q value, the first two X/Q values should be compared and the higher value selected.

This value is then compared with the third X/Q value and the lower of those two is the appropriate X/Q value. The appropriate X/Q is 5.3 9E-06 sec/mr 3 for default meteorological conditions by the methodology recommended in Regulatory Guide 1.145.This calculated value is very similar to the ODCM highest average value of 5.3E-06 sec/n 3 which was not selected for use. Additionally, the value shown in the STAMPEDE output file at one mile is 1.032E-05 sec/m 3.This suggests that STAMPEDE uses the same methodology and simply selects the largest atmospheric dispersion value to remain conservative.

This methodology will be replicated and 1.07E-05 will be used as the X/Q.:7.-.-2 N lidefInventory......

............

7 7 .... ...................

..As previously stated, a gap inventory is appropriate for this problem. The gap inventory is taken from TGX/THX 3-1 (Reference 5.4) which is used as the source term for STAMPEDE inventories.

The concentrations were then normalized so they could be scaled to the varying emergency classifications.

The values for the normalized inventory can be found in Table 4.2.7.1.3 Dose Conversion Factors As stated in NEI99-01 (Reference 5.14) developer notes, the purpose of dose projections is to check if the Environmental Protection Agencies Protective Action Guidelines (EPA PAGs) have been exceeded.

The dose conversion factors provided by the EPA in EPA 400R92001 are used. These dose conversion factors account for external plume exposure, inhalation -from the plume, and external exposure from deposition and are listed Tables-+.,- WIiU '.J3,' LW IW1 )1J~11 LU 1 -J.- ,-, 3-. HI nu"-k r'VU1\.YeUU f

l 3.4.)..The EPA does not provide a dose conversion factor for Kr-83m. This nuclide contributes 1.33% of the inventory activity.

The lack of this nuclide's contribution to the final dose will not significantly affect the outcome.7.1.4 Decay Time One hour of decay is incorporated to the monitor response due to the release initiating one hour after reactor shutdown.

Decay is also incorporated for the duration of the migration time. The total decay time is one hour plus the reciprocal of wind speed, or 1.07575 hours.

Radiological Release Thresholds CALC. NO. STPNOC013-CALC-002 EN ER CO N for Emergency Action Levels "RV.I fta ,i-Ey P dyAttachment 1 PAGE NO. 22 of 49 (7.1.5 Dose Calculations The dose rate at the site boundary is calculated using Equation 7.1.5.1.X .. .n1.07575.D = Ci

• 0..5 'DCF Equation 7.1.5.1 Where' dose rate per hour at the site ounidary x atmospheric dispersion coefficient as calculated in section 7.1.1'.F u"nit vent flow iate ..C., conc-"d eetration of nuclidi iat the time of shuitdow t 0575" the total decay time of interest from section 7.1.4-Tti.=- the half-life.

of ri.uclide I'DCFt.t the dpse. 6nversion factor for nuclide i listeed-in tables 4.4 and.."4.5 1: The total iaiiclides is varied to find the.dose rate of interest.Beginming with an arbitrary release concentration of 4Ci/cd."the dose rate is calculated.

Since the dose is linearly correlated t0o coicentrat ion, thejrelea, econcentration may be scaled to find the dos erate ofinterest.

The Alert EAL[s 10mrf" i TEDE or 50 mrnm Thyroid CDE. Using the above method to calculate TEDE with the appropriate conversion factors, a timiflng release rate of.:.2.33,E+-06 the unit ventresultsin 5.7.mrrm TEni .,.Ui ng the calculated release rate tofmdyd T D iththe propriat ecnversion factors, the same release resulý iin a 50 m.rem Lhyroid .CDE at the site boundar.;

Thus, 2.33E+06 gCi/sec is the limiting rlease rate based on the.50 mirem Thyroid C1'l EAL" fiitiating condition.

The limiting release fate threshold values for the Site Area Emergency and General Emergencies are multiples of 10 and 100 of the Alert release rate threshold value.These calculations can be found in Attachment 2.7.1.6 Monitor Response.The unit vent noble gas monitor is: calibrated tO Xe-133. Monitor efficiencies relative to Xe133 by nuclide are listed in ODCM TableB3-2.

To findthe monitor reading associated with each limiting release, the noble gas concentrationsmust be multiplied by the monitor response and sumned. Table 4.6 shows the indicated response of the unit C>

Radiological Release Thresholds CALC. NO. STPNOC013-CALC-002 E N E R C O N for Emergency Action Levels REV. I ,xcefie-Ewvy ptofta Fy, day Attachment I PAGE NO. 23 of 49 vent noble gas monitor by nuclide and Equation 7.1.5.1 shows how the monitor response was calculated.

n Monitor Response = C 1

• Re 1 Equation 7.1.5.1 Where C 1 = concentration of nuclide i Re 1 = monitor response to nuclide i (GCi/cc)x.

1 3 3 equivalent In the case of an Alert, the 2.33E+06 iCi/sec release rate will read as 1.57E+06[LCi/sec on the monitor. Kr-83rn does not have an indicated monitor response coefficient.

Because Kr-83m is only 1.34% of the noble gases and does not contribute to the dose calculation, its exclusion is acceptable.

This again is a linear correlation and the SAE and GE scale by factors of 10 and 100 respectively.

These calculations can be found in Attachment 2.7.2 Main Steam Line Monitors 7.2.1 X/Q Since the atmospheric dispersion is independent of nuclide inventory or release rate and the close proximity of the releases, the X/Q value will be the same for a main steam line release as it is for a unit vent release. This assumption is also taken by STAMPEDE and outline in Assumption 6.3.7.2.2 Nuclide Inventory Per- UEKRPk1 ZVT, POTif the release path is Me main steam line with a steam generator tube rupture, the nuclide inventory should be 100% noble gas and 0.2% of the iodine from the reactor coolant.The secondary steam concentration for noble gases and iodine after a steam generator.tube rupture are taken from TGX/THX 3-1 (Reference 5.4). Values for the reactor coolant inventory are listed in table 4.3. All of the noble gases are used and the iodine concentration from the coolant inventory is scaled to total 0.2% of iodine in the total coolant inventory.

These inventories are then normalized to one. These values are listed in Table 4.3.7.2.3 Dose Conversion Factors The dose conversion factors used are found in Tables 4.4 and 4.5, taken from tables 5-1, 5-2 in EPA 400R92001.

Radiological Release Thresholds CALC. NO. STPNOC013-CALC-002 E N E R C .N for Emergency Action Levels m 7.2.4DecaS'Time Attachment 1 PAGE NO. 24 of 49 1J 7.2.4 Djecay Time.One hour of decay is incorporated to the monitor response due to the release initiating one hour after reactor shutdown.

Decay is also incorporated for the duration of the migration time. The total decay time is one hour plus the reciprocal of wind speed, or 1.07575 hours.7.2.5 Dose Calculations Equatfio 7.1.5.1 applies to the release from the main steam lines. The main steamn line oflw rate is used instead of the unit vent flow rate for the value: F. The main steam line flow rate was calculated in Equation 7.1.2.2 of the STAMPEDE CALCULATIONS section of this document as 2.79E+06 cce/e.The Alert EAL threshold is 10 m'rei TEDE or 50 mrem Thyroid CDE at the site bounda'ry (Table 42); Using the .ethod in Equation 7.1 .5 to' calculate TEDE with the ,apropriate conyrsiOn facqors, a c6ncentratio'n'aat time of shutdown of 4.10 .tCi/cc would' egulti i0.89 mrei TD"E 4a1he iteiboidaim y if the §team line PORV was open for an' hour. ,Usig tihesae steam line concentration to5ca:6iclate.'Thr~id CDE results in 50 mrem Thyroid CDE at the site bobndaryi.

--..The steam line-.q.cef~itrations f.for the"Site Area-Emergency and --G General Emergencies are of 10 ind 100o pth6 Alert Sinice the correlation between release concentration and dose is linear, valiues for the: ste'am line concentration at time of shutdown are 41.0 and; 4:10. Ci/cc for the SAE and GE respectively.

Both are also( limited by Thyroid CDE" ..These calculations can be found in Attachmiient

2. .7.2.6 MonitoriResponse.

Becauis the main steam line monitoris adjacent to the tain stearA line, significant shielding takes place between th-o Sce'andmonitor.STP calculation NC-9011 Revision 2 calculates a conversion factor for the'mai steam lines~for a noble gas inventory which is incorporated into'the monitor readout. No monitor response needs to be calculated.

The corncentration of the main steam line one hour after shutdown given a concentration of4.i10 Cicc atttie of"shutdown isi3 .0 pCi/ccTfhis calculdiioniis also foufid in Attachment

2. Addition y, t hit6r feAdings fo6 the SAE and GE one hour after shutdown are 39.0 and 390 pCi/cc respectively.

These values are the thresholds for the main steam line monitor.-.I CALO. NO. STPNOCO I 3-CALC-002 Radiological Release Thresholds foErN EREthgency Action Levels$rcehnc--S'r/$t4t "Attachment 3 PAGE NO. 32 of 49 DRILL' 'L7 ~ ~:r luomi D LL air .ME .A; ..Afr.t L~nrd~dDunr I mtaas'Ed~nS**IDSmSS 120117W3MUM Nq m 34a3 .War" (u& D I Et-t 105B04 n 31UUB tll: 25EMS fl-S .: pmi 4I..*d:' : i-lid ImR Ce-Ut : :. ZASWi: ,': : Xe-n- GE400h KuU:2f Im ,1?AOI 3-M.46 FM R2ological Relese Thresholds CALC.NO.STPNOCO13-CAL-002 0 E NE CO N for Emergency Action Levels REV. I Attachment 2 PAGE NO. 25 of 49 Table A2-1: Unusual Event Emergency Calculations

..,I 1 .40E,+05 I 2.79E+06 1. 5.OOE-02 Table A2-2: Input Values for Calculations Radiological Release Thresholds CALC NO. STPNOC013-CALC-002 OM m. for Emergency Action Levels REV-.Attachment 2 PAGE NO. 26 of 49 Table A2-3: Ca.culations for Boundary Concentrations and TEDE dose due to Unit Vent Release 1-132 1-133 1-134 1-135 Kr-83m Kr-85m Kr-85 Kr-87 Kr-88 Kr-89 Xe-13 lm Xe-133m Xe-133 Xe-135m Xe-135 Xe-137 Xe-138 1.50,E+0.5 2.20E+05 2..40E+05 2-OOE+05 1 .30E+06 2.90E+06 3 .70E+05 5.504,06 7.80E+06, 9.50E+06 1.1OE+05 6.80E+05 2.20E+07 4.20E+06 5.50E+06 1 .90E+07 1 .80E4+/-07 1.53E-03, 2.125H-03.

2.05E-03 1 .33E-02 2.97E-02 3.78E-03 5.62E-02 7.90E-02 9.72E-02 1. 12E-'03 6.95E-03a 2.25E-61.4.30E-02 5.62E-02~1 .94E-01 1 .84&-01 3.9 I 1 2 I2 5, 1: 1..* .4, E-5$5F-05 64E-05 6&-05 0E-04, 533E-04 33E-05 39E-63 97E-03 40E70' ..76E-05 71F,04, 5-5E-03 06E-03 39E-03 79E. 03, s4E-03 1.OIE-03 1.QIE-03'1-0 1B-03 1.0 E- .03 1.0 IE-03, 1 .01E-03 1.01B-031 I .O1E703 1.01E-03ý1.O1E-03 1.OIE-03 1.0 1E-0ý3 3,84-&-08, 2.38E+00 5.61E.-08 2.03E*10-6.1 IR-0 9 -.77E-0 1-5.II-T& 6.6 IE40.0 3.3 1E-07 11.83E+tO0

.7.40EM7 ý4.48 .E+00 9.42E-09.

9.4oE+04 1.40E-06 1 .27E+00O 1.99t-06 2.-84E+00 2.42E-06 5.1OE.-02.ý

-2.79E-Q08 2.83E+02--1.73E-07 5.42E4-01.

5S61E-06 1.27E+02 1.07E-ý06

-2.66E-01 I AOBE-06 9.09E+00~4].5ER-06 638E-02 4.59E-06 .23E 2.79E-08.S.40E-08-2.f61R08 2.21F-07 6.27]E-07 9.42E-08 7.79E-07 1.53E-06 2.78B409 1.711E 07 6.09E-08 1L29E-06 4.06F,,11 1 .95E-07, I.50E+034-3.1QE+03.

8"10E+03 9.30E+0 1 1.30E-I00 5.10E+02 1 .30E+03 1.20E+03 2.50F,+02 1.40E+02 1.104E+02 7.20E-I02 5.30E+I04 4-90E+03 1 .50E+04.1.37E-04 8.11E-04 ,09E-05 3.70E-04 0.OOE+0-0 5.83E-05 1.22E-07 3.97E-04 1.99E-03 1.30E-09 1.36E-07 2.90E-06 1.11 E-04 1.52E-05 1.81E-04 4.47E-09 1.40E-04 10 E E O for Emergency Action REV. I Attachment 2 , PAGE NO. 27 of 49 GS-134 3.70E+01 3.78E-07 93 -09 L.O1E-03 9.42E-12 i.80E+04 9.42E-12 6.30E+04 5.93E-07 Cs-137 2.90E+01 2.97E-07 7.3 E-09 1.01P-03 7.40E-12 2.60E+05 7.40E-12 4.10E+04 3.03E-07 Te132 4.80E+00 4.91E-08 1.2 E-09 1.OIE-03 1.22E-12 7.79E+01 1.21F-12 1.20E+04 1.45E-08 Mo99 1.22E+01 1.25E-07 3.0 E-09 1.01E-03 3.11E-12 6.62E+01 3.08E-12 5-20E+03 1.60E-08 Rul03 8.80E-03 9.00E-11 2. E-12 1.O1E-03 2.24E-15 9.44E+02 2.24E-15 1.30E+04 2.91E-I I Ru106 2.90E-03 2.97E-1I 7.31-13 1.01E-03 7.40E-16 8.84E+03 7.40E-16 5.70E+05 4.22E-10 Zr95 Lal40 Ce144 Ce- 141 Sr89 Sr9o 1. 1 OF,-02 1.90E-02 7.40E-03 1.O0E-02 6.40E-02 3.20E-03 1.12E-10 1.94E-10 7.57E-1 1 1.02E-10 6.55E-10 3.27E,-1 1 2.7 4.7, 1.8'2.5S 1.6: 8.0" E-12 E-12 E-12 E-12 E-1 I E- 13 1.OIE-03 1.0IE-03 1.01E-03 1.01E-03 1.OIE-03 1.O1E-03 2.79E-15 4.83E-15 1.89E-15 2.54E-15 1.63E-14 8.15E-16'1 .55E+03 4.033E+0 1 ,6.82E+03 7.77E+02:1.21E+03 2.50E-405 2.79E- 15 4.75E-15 1.89E-15 2.54E-15 1.63E-14 8.15E-16 3.20E+04 l.10E+04 4.50E+-05 1.10OE+04 5.00E+04 1 .60E+06 8.93E-11 5.22E- I1 8.49E-10 2.79E-1 1 8.16E-10 1.30E-09 T6~d F~D~ Dose -, 5.~77E-O3

-~Radiological Release Thresholds CALC. NO. S1Th~o13-CALCOO02 0 ENE C.O N for Emergency Action Levels REV.°I Attachment 2 PAGE NO. 28 of 49 Table A2-4: Thyroid Dose Calculation for Unit Vent Release 1-131 1-132 1-133 1-134 1-135 2378E-08 2.79E -08 5.40E-08 2.61E-09 4.56E-108 I '30E4,06 7.70E-f 7 3 220E+05 1.30E+0O3-2QvLn 3.6IE,-02 2.15E-04 1.19E-02 3.39E-05.1 '7)D n V27 Table A2 Unit Vent Monitor Response to Nuclide Inventory Kr-85m Kr-85 Kr-87 Kr-88 Kr-89 Xe-131m Xe-133m Xe-133 Xe-135m Xe-135 Xe-137 Xe-138 U.33E-04 P33E-05..39&-03 ,97B:03 ,;40E-03 ,.76Eý-05.71B-04.;.55F-03.06E-03.39E-03-.79&-03-.54E-03 4.48E4-00 9.0E+-04.1.2"E00 2.84E+01O 5.10F,02 2.93E+02'5.42E+01..

1.2-7E402'.

2.60E-0 1 9.08E-+00 6.38E-02 2.3 6E-0 1 2.25,F-04 6.28E-04, 1.9 9.33E-05 2.4 8.03E-04 2.8 1.54E-03 2:3 3.00E-09 Z.8 2.76E-05 0.015 1.69E-04 0.14 5.52E-03 1 7.38E-05 0.042.1.28E-03 2.5 9.15E-08 2.8 2.41E-04 2.8 Monitor Reading: 0.OOE+0O 1.1 9E-013, 2.24F,04.2.25E-03 3.55E-03 8AO0E-09 4.13E-07 2.37E-05 5:S2E-,03ý 3

• IE-06 3.21E-03 2.56E-07 6.74E-04 KUM~c) (uCi/sec)

Radiological Release Thresholds CL.NO. STPNOCO 13-CAL(>002 SE N ER CO0 N for Emergency Action Levels REV. I awo-mPt&~dard Attachment 2 PAGE NO. 29 of 49 Table A2-6: Input for Main Steam Line Release Calculation I 1-1-1 1-132 1-133 1-134 1-135 Xe-131m Xe-133 Xe-133m Xe-135 Xe-135m Xe-137 Xe-138 Kr-83m Kr-85 Kr-85m Kr-87 Kr-88 Kr-89 0. 1Url-UL 8.61IE-02 I .OOE-01 1 .86E-02 2.73E-0 1 2.80E+00 2.40E+02 4.20E+00 7.60E+00 4.OOE-Ol 1.60E-01 5.80E-01 3.70E-01 7.60E+00 1 .50E+00 9.80E-0l 2.80E+00 8.40E-02 3.19E-04 3 .72E-04 6.92E-05 1.01E-03 1.04E-02 8.90E-01 1.56E-02 2.82E-02 1.48E-03 5.93E-04 2.15E-03 1.37E-03 2.82E-02 5.56E-03 3.63E-03 1.04E-02 3.12E-04 Y.L I /-U't 1.31E-03 1.53E-0.3 2.84E-04 4.14E-03 4.26E-02 3.65E+00 6.40E-02 1.16E-01 6.07E-03 2.43E-03 8.82E-03 5.62E-03 1.16E-01 2-28E-02 1.49E-02 4.26E-02 i .28E-03 1L.Y035f-1rJV 2-9853E-05 2.9853E-05 2.9853E-05 2.9853E-05 2.9853E,05 2.9853E-05 2.9853E-05 2.9853F,05 2.9853E-05 2.9853E-05 2.9853E-05 2.9853E-05 2.9853E-05 2.9853E-05 2-9853E-05 2.9853E,05 2.9853E-05

/1. / IrD-Uo 3 .90E-'08 4.55EI 08 8.47E-109 1 .24E-07 1 .27E!06 I .O9Eý.04 1 .91E-06 3.4513,06 1.8 1E-07 7.26E.-08 2.63E7-07 1.68E-rO7 3.45E-166 6.81E-07 4.44E-07 1 .27E,406 3.82E-08 I .Y.D"Jh., 2.3 8E+00 2.03E+01 8.77E-01 6.61E+00 2.83E+02 5.42E+01 1.27E+02 2.60E-01 9.08E+00 6.38E-02 236E-01 1.83E+00 4.48E+00 9.40E+04 1-27E+00 2.84E+00 5.1 OE-02 Z. /I OV--o 2.85E-08 4.39E-08 3.62E-09 1.1 OE-07 1.27E-06 1.07E-04 1.90E-06 1.96E-07 1.67E-07 6.10E-13 1.12E-08 1.12E-07 2.92E-06 6.81 E-07 2.47E-07 9.79F-07 1.71E-14 4.90E+03 1.50E+04 3.101E+03 8.10E+03 4.90E+00 2.00E+01 1.70E+01 1.40E+02 2.50E+02 IAOE+02 7.20E+02 I.L+OZ-UJ 1.401-04 6.58E-04 1.12E-05 8.95E-04 6.22E-06 2.15E-03 3.23E-05 2.75E-05 4.17E-05 8.53)E-11 8.04E-06 O.OOE+00 3.80E-06 6.33E-05 1.26E-04 2_OSE- I1 1.30E+00 9.30E+01 5.10E+02 1.30E+03 1.20E+03 1 .r I~ose, G. 03z-O ,*Release Constant = X/Q

• duration

• release rath 1-1311 1-J32 1! 133 11&5E-08-4.39E-08 3.6.2E-09'7.70E+ý0-3 2.20E+0.5 1 .30E+03.3.58E-02: 2.20E-04 9.66E-03~4.71F,06 4.20B-03 Radiological Release Thresholds CALC. NO. STPNOCOI 3-CALC-002 0 E N E R C O N for Emergency Action Levels REV. 1 Mx*-Y dyAttachment 2 PAGE NO, 31 of 49 Table A2-9: Main Steam Line Reading at Release 1-lij 1-132 1-133 1-134 1-135 Xe-131m Xe-133 Xe-133m Xe-135 Xe-135m Xe-137 Xe-138 Kr-83rp Kr-85 Kr-85m Kr-87 Kr-88 Y.Z /ts-U4 1.31E-03 1.53E-03 2.84E-04 4.14E-03 4.26E-02 3.65E+00 6.40E-02 1.16E-01 6.07E-03 2.43E-03 8.82E-03 5.62E-03 1.16E-01 2.28E-02 1.49E-02 4.26E-02 2.38E+00 2.03E+01 8.77E,01 6.61E+00 2.83E+02 5.42E+01 1.27E+02 2.60E-01 9.08E+00 6.38E-02 2.36E-01 1.83E+00 4.48E+00 9.40E+04 1.27E+00 2.84E+00 S.IOE-02 9.23E-04 9.77E-04 1,47E-03 1.29E-04 3.73E-03 4.25E-02 3.60E+00 6.36E-02 8.04E-03 5.62E-03 4.65E-08 4,67E-04 3.85E-03 9.90E-02 2.28E-02 8.62E-03 3.34E-02 Radiological Release Thresholds CALC. NO. STPNOC0I3-CALC-002 EM.. N for Emergency Action Levels-Attachment 3 PAGE NO. 32 of 49 rabisiftke Alii; 1ilIm3 Ni KE~S~ra~ 1I~4@~uG~ur vNOMZ4TO410 a .4f Kr43t 13E0"I Xe-33M UWm40OB Yw-uSB1 9-B&IM 3120MII 3&3,W- 3.03-T -2 13E.X~Z[4-i .3im .tm (MMOUM33~AP?

C CALC. NO. STPNOCO13-CALC-002 Radiological Release Thresholds EN ER CO N for Emergency Action Levels I xce*-Erypr.

Ervyday. Attachment 3 PAGE NO. 33 of 49 DRILL-32=01 AV0 DRJLL Dubaime M7114OI 15:M Ukw MM~E UMMIk uAjat UiStWj (miles)05.7.5 10D 2.0 520 7.3 20D (mistu&s l.0 7.5 IA0D MaD 023 OL45 1:31 4~iQ Vabu~cJ~2.Ui~.O05 im~3.IS1R-007 U~007 9.11l.0'3.ISIE.O0 7.373H,07 3245ECO1 1441E-007 amo XmshuknloBody owl 0.003 0.001 Row0 0"0 rdzoeD in ruiv C Q'07~~0.04 (Sam) OflE 0M00 0.001 0M0 CamO 0237 0.002 am~amo 12=0t33142*28 CALC. NO, STPNOC013-CALC-002 Radiological Release Thresholds REV. I E N. ER C 0 N for Emergency Action Levels. W.t$icm-4n rymJed ýily dV. Attachment 3 PAGE NO. 34 of49 C DRILL* ....STAMPLDMResuiltsIto--S -...RESMLTS... W InuI,. Aedti 132 ...Jh ..'.:: a ':+ " ....?P DRILL Re..a... .te: LL1 .* ..ui....IOffh&t3 neurajMane (reier CUE D .051i H0ICL O . Emjected durtionoafudezeet Kohour, A Gener~all Emergency eY1qnires a -Pro'tective Action Recommemilation EVACUATE ZONE() I SHELTER 19 kLAE ZONE<SI 2 AFFECTED DOWNWIND SECTO1I& Reý A,B All Reauliag Zones Go fiidori And Monif or EAS mlic Station.Based outalDose Rote Pr'etaiearef 3 inamter/b (inxsiou Whole Blody No ble Gas Gamma) at the S ite B oimduzyQ( Mile) ftr 15 ine or longer &he Enwmeeny Ch1.9kfeelio I itatig C4ondition R.1I (ALERT.) has been met A REVIEWED BY: dMnsxerRmdoloAk-Dirnetor.... l2/20153:24&44PH DaitddTie 1211712013 3.:24:28PM CALO. NO.

Radiological Release Thresholds EN ERCO N for Emergency Action Levels UV. I, Attachment 3 PAGE NO. 35 of 49 DRILL STAMPEDEUser

%~ppfied InformatiouDR L DRL sT ..&F-d' !Mn7M3. .... DRML Im~hfflm:

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CALC. NO. STPNOCO I3-CALC-002 Radiological Release Thresholds ENE R CO N for Emergency Action Levels REV. I Attachment 3 PAGE NO. 37 of 49 DRILL ssTAMPwE03 Weut IngformIofi2 D.NRILL-alcuadIiens CID ylefed RESULTS.V~uDireetizma ISO Methodd bPmrfrial STA1M2D Rdewaaa Ih .19H+Dwu0&C~M I me ml~es ~ ml~es 10miles TEDE 0407 D.003 M.00 DAD0 CDE 0.050 0.017 0.004 1.001 Prouc4inmn fr~a~e~1.0 bour A General Emergency Requmires a Protective Action Recommendation EVACUATE~

ZO?4EMS: 1 SHELTER IN PLACE ZONE(S): 2 AFFECTIDDEOWNWTND SEC-TORSt

-, A. B .....-.All IPernining Zones Go. Indoors And Manitur EAS Radio Station 13 ased on a Dose Rate Projection of > 3 mremuhr (Immersion.

Whole Body Noble Gas Gaumn) at the Site Bowrdary (I Mile) fci 15 mimites or longer flit Emergency C aificfimi~n Initiating ConditiouRAI (ALERT).Itusbeeumnet.

PRFGMRD BY.121l180137355:4AM DatelThne R --- ----- ---- H w i -IT2I EW213 7:54-42AM CALC. NO. STPNOCO 13-CALC-002 Radiological Release Thresholds E E-R C0 N for Emergency Action Levels"fn-tý ..qf1Wt PyOyd.. Attachment 3 PAGE NO. 38 of49 DRI2___L " " .2_tnm_' ~f.. _DRILL ihwflum 12t1~X2Ot3 1525 , twlknr 1frk~$atSihAao Coals Ike'. It L MatwdwatnllhtYnwtz,*atfru1~utdaUr.

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A ee~ mrency Requires at Prateletive A~tio Reomndatio EVACUATE ZONW(). 1 SHBLTER IN PLACE ZMIE(S). 2 AFFE.CTED-DOWNWJNI RET1S , A, H All Rem "ft Z w hs x~ In t A nd MwidtoeEMS Ib dia 'ýAti M: R~em&Becy CJsi[.fcatwn IitfialngConiitwicnS1 (SIlE.AIEA EMERGENCY) batbeenm aet Al-PERFO0AIDUY DN~mt I -~QO1325:23 FM 12FI712013 3:25.11 FM GALC. NO. STPNOCO I3-CALC-002 Radiological Release Thresholds RAEV. IO. SPNO __3-CA _-00 E N E R C O N for Emergency Action Levels REV. I cehn-Mq t Attachment 3 PAGE NO. 41 of 49 DRILL STAMPEDE User Suppled lorsato DRIL PaWaLme 1217f0 1523 CMnunfic lkW~aMsc StsmmbaSfleAmn Qa~dlwmtuubtr31 ni&rs Gmdcrkuivi=dhrw 12 &gM1ree Mam-sMIkctuifbm-LODltun Nrbdae: ltllf 3.÷1 NovdiakW tfnAre ride:o: 1* 4as Cumlclte&OMZC4llSrdmsrte:

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_ay. Attachment 3 PAGE NO. 43 of 49 DRILL STAWIEDEResults Infonntftrn DRILL CuiazCIDMn.

1-tetd RESULTS Wffiad: ot~rajecdca WAxd~kI*-4 132=Ulx Reeene 12EO STAMPI1DE idretm19 Off ft Do" roiectiou (r'em)i, I nol 1 MILOS Bmfles 1 Mil"s IEDE 0.02 25 0,006 ~ 0M CM 0.506 WI5 0.0)41 0.013 Pnwjected dumfiofreleasr.

1.0houni A General Emergency Requiires a Protective Action Recarmvnudatioin EVACUATE ZONE(S) I SHELTER IN PLACE ZONE(S: 2*AFFECTED DOAVMVIN~D SECTORS. R, A, B -All Remmining Zowe Go Mdoone And MornitorEAS Radio Station fse~Based on a Site floandazy (I M~'ile) Dose Prjec~tion

> 0. 1 rem TEDE imdlor DJ5 xm Thiyroid CD-E the Emergency Cisanifration Iuitiatin-CondtioniRSI (SITE AREA EMERGENCIV) bni been met.I PEKTORME B~r REVIMVD BY, 1IMi7013 519-10D PM DatsThoe a.1201/2013 3:203pM CALC. NO. STPNOC013-CALC-002 Radiological Release Thresholds REV. I 0 EN ERCO for Emergency Action Levels V.

yptrjda: Attachment 3 PAGE NO. 44 of 49 (nun y STAMVEDE User Supj1Ii~I Tatformation ThD b~tsnral Dat Inf Tka-sdactuS61flifika Mubmd~3it:ad~r

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CALC. NO. STPNOCOI 3-CALC-002 Radiological Release Thresholds aa'E NE RC0 N for Emergency Action Levels REV. I Evry doX Attachment 3 PAGE NO. 45 of 49 DRILL STAMEDE Lobs ImIormahoa DR]LL DIzdrumt IYIfW 1.5- UffNam: Tht Wok,-nd 4 4 DhlnMcu (VaiIes)aks 1.0 2.0 5.0 73 10.20)Pluma lkzd !Mao 0zuunMiwaleul U2 ams ODD 0:23 05"4 OAS 151 CULQ Vain&(ssrM'3 2069M 1OMM?7373EM0 2.44IB00 93107300.0 io 5.0.7.5 DiIun~ca 05 ID 10 510 7-3 MD 20)nM~ris. Val& OMITf 0,3322 tin 0M01 0.010 0IM73 Yam~rnla W&At Body nobsb go, grms (raw)till?0"01 0.000 0.0Mz MW CM.xeerwi+ ftrww. 2vIdyr.. umt" hrlY -.""15 0)01 0(210 0)5 ami CW17 OW(1 Snm 1,762 4131 U314 0135 10M0 WE baie CURI 0M60 am" C.M7I3M" Szo 1.762 0.411 0-214 0-135 ame 12(ilflflh3flt23PM

.4 CALC. NO. STPNOC013-CALC-002 Radiological Release Thresholds EN E RCO N for Emergency Action Levels _ _lV._ _ _£rseEveiypjV

... ...Attachment 3 PAGE NO. 46 of 49 C DRILL STAMUME Result Intmtiion DRIL* ......,e.' C o ma '. .STAIQPfDE TEDE I0 1(LS1U4S W~nd~L~d .232mib WDW~ok~ a0s'02. M411 ~ nie lOg.~1IleB 0.017 DM5 A Geuerai iiaýxgency Ac.eauires a Pratertive'Actwin R. e. ommiednition EVAICUATEZONE(Sy-li1 SHELTER W PLACE ZON'E(S>.

6,11.AF~FECTED DiOWN WN ECTORS. R~, A. R'AJI Remainig Zom. (io hnDI A~6Mwj3mdrEAIS

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CALO. NO. 5TPNOCO1I3-CALC-002 Radiological Release Thresholds CAL_, NO. STPNOC013-CALC-002 I E IN E R C O N for Emergency Action Levels Rv 1 frcefr,-rwryn Evet Fryda Attachment 3 PAGE NO. 47 of 49 DRILL 'TZD Ur fd ormatio DRILL Da c12T 1530 Cm-M& I U", M n. StnMaUSUGMe" ThrchM ilnfmint Mtewardnu~d*Fht rapt: --- -- --a"----a---td sb~ry*" m" Shempbycist

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M102-MD .0sav NOBIEGAS Macida uCIuw N 123Wme PAMKUEATE Macdi* .01/n 3t-S: ia-sw Kr-S.K8r-iD: Xfl-MmXT-Ut Xe-Ut 114940W6 3AM&+00 SAMMf~l Z n+C 413E41M1 1MMhw7 rLW:.-13M 1.-135: 424B01 IMMOM0 Ce-IC1: lnk D.0.3.M0 DOXff4 SrJY-5t Sr-W: oa0,+12n7.20133:).:SOwt CALC. NO. STPNOC013-CALC-002 Radiological Release Thresholds REV. I E N E R C 0 N for Emergency Action Levels. 1.7&n- A* Mrydao Attachment 3 PAGE NO. 48 of 49 C STA.. rlm.onn.iC DRILL DsWIr 32fl7D013 1510 Uflmet Klamhmn..... ......, ....* * ., , ........I .i d ýj ý d ýIp PM -I., -., -b- , -, .:1. .. .j :, I ;p IM-ý 3 M jT-v,7 -7 :' ý' , ' , ! --ý : --ý, ý I I : -, -d ; -, ::;Ehhmm 75 20D 05 10.0 062 1434F5 045 1725524 24MM01 44236I~e~

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STAMIPEDE RdesI~anFrrt 1.0E+009uQihc 0ffEZIeCD ePrqjeLtiwi(rLL=)i 7M DJA 0154 CLO63 am2 CUE ¶1.47 M0(7 0.134 Pwjedad Jti~aofr4e~ame:

1.0 hour0 days <br />0 hours <br />0 weeks <br />0 months <br /> A General Emergency Re quires a Prote~ctive Actiou Recainmendation EVACUATE ZWNE(): 1, 2 SHELTER IN PLACE ZONE(S: 6, 11 AffECTED DOWNWIN SECTORS: -R, A, B All Remainiing Zeone Go bn$oms And Monitor EAS Radio Statcoa Based ani a Site Boandary (I Wie) Dose P oection ý I remn TEDE andior. rtem Thyroid CDE the Emergeur Clasifrsienfi nitintin 5 Condition RGI (GENEAL EMERGENCT) has been met pmwnIU~IE By: REVIEW BY, 12012f013 130:V48 PM Thitetrhn______________________________________________

1~12417#2013 3:.313PM STPEGS UFSAR The particulate channel is used as part of the Reactor Coolant Pressure Boundary (RCPB) leakage detection system. The sensitivity and, response time of this pare of the leakage detection system, which is used for monitoring unidentified leakage to the Containment, are sufficient to detect an increase in leakage rate of the equivalent of one .gal/nin withfi one hour. Elements of this monitor, Including -he indicator mounted in the RMS CR cabinot, are designed and qualified to remain finotional following.

a Safe Shutdown Earthquake (SSE),. in -omplianoe with RO 1,45.. Further' i information on the RCPB leakage detection system is presented in Section 5.2.5, 11.5.2.3.3 Unit Vent Monitor: The unit vent monitor samples the plant vent stack prier {o discharge t6 'he environment and monitor for particulates, iodine, and noble gases.The unit vent particulate.

and iodine monitor draws representative air samples from the plant vent stack. via isoldnetic nozzles in the stack, and -directs them through a moving filter paper monitored by a shielded beta-sensitive scin' illation detector.

The sample stream then passes through a charcoal collector; where collected iodine is monitored by a shielded gamma-sensitive sointiilatibn detector, The sample is then returned to the vent s.tack.A separate wide-range gas monitor is p'rovided for the unit vent. The monitor, has two isokdnetic nozzles for sampling.during both nonmal and accident conditions.

The stack samples pass first through a sample conditioning unit which filters particulates and Iodine and may' be used to. take grab, samples, The samples then pass through the shielded detector assembly,, which uses throe detectors to cover the complete range required.

The low range detector uses a beta-sensitive plastic sointillatorlsphotomultiplier (PM) tube. The mid-range and high-range detectors use cadmium*'.., ~elluride (CdTe), chlorine-doped, solldstate sensors, This wide-range gas monitor satisfies' the requirements of NUREO-0737, Item 11.F.,1 for provisions fbr sampling plant effluents for iodines and battioulates and f6r noble gas effluents from the plant vent.11.5.2.3.4 Contol Room Eiotrical Auxiliary Buildina Ventljation Monitors:

The CR/EAB ventilation monitors are Class 1B monitors which continuously assess the Intake air to the CR for indication of abnbrmal airborne radioactivity concentration.

Bach monitor assembly is powered from a seprate. electrical power source, In the event of high radiation CR emergency ventilation operation is. initiated .Seotion 7.3.2). Failure of a monitor Is alarmed in the CR.EB&ch monitor assembly Is oomp-rsed of a recfrculation-pump.,3etagSensi1itlv ,oiitifit idn "....fourMpi lead shielding, check source, stainiess stelo sample gas receiving' chamber, and associated electronics, 11.5.2.3.5 Condenser Vacuum Pump Monitor: Gaseous samples are drawn through an off--lne-system by-a-pump-from the-disoharge-of-the-vd6iuum-pump-exhaust-header-of-the-condenser.

-_This ohamel monitoxs the.gaseous sample for radioactivity which would be indicative of an SO tube Jleak, allowing reactor ooolaut to enter the secondary side fluid; this monitor complements the SQBD'itonitors hi indication of a SO tubo leak.' The gasebus radioactivity levels are monitored.by a fingle detector in a manner similar to the unit vent wide range 'gas monitor.'1 11.5-2.3.6 Snent Fuel Pool Echaust Monitors:.

The SFPE monitors are Class lB and ', identical to the CRIEAB ventilation monitors described in Section 1.1.5.2.3,4 except that they sample the exhaust from the FHB, In the event of high radiation the monitors initiate emergency operation S1l.5-i.1 Revision 14 STPEGS UFSAR (" 11.5.215.1 Gaseous Waste Processing System Inlet Monitor: The GWPS inlet monitor I employs a gamma (Nal crystal) scontillator/photomntflplier tube combination to measure the radioactivity level of the waste gases entering the G(WPS-. The monitor is used In conjunction with the GWPS discharge monitor to measure overall effectiveness of the GWPS.11.5,.25.2 GWPS Discharge Monitor: This monitor is similar to the GWPS inlet monitor and is installed upstream of the OWPS discharge valve. Upon detection of high radioactivity or monitor failure, the GWPS discharge valve, FV-4671, is. automatically closed, 11.5.15.3 Main Steam Lihe Monitors:

Each.MS line is-monitoredby .anATL monitor consisting of a Geiger Mueller (OM) tube detector and an ionchamber detector with overlapping ranges. The detectorsare shielded by 3 fn, of lead.The monitors are designed to moziitor gross gamma activity in the steam line. and provide a basis for detentAning possible atmospheric releases from the MS power-operated relief valve (PORV), SO 04 safety valves, and/or auxiliary feedwater pump turbine, The monitors provide a dose rate range equivalent to 1.071 to 103 ikCYm 3 xenon-131.

Based upon core inventory,.

the ratio of xenon-133 to other nuclides in the fuel can be dotemined, In order to obtain the abov6concentrations of xenon. 133 in the main, steam line, a large primary-to-seoondary leak must be present coincident with a large amount of fael failure. The presence of xenon-133 (. indicates other radioactive isotopes ame pmesent.Using the relative ratios ofisotopes present in the MS line, -a computer model for determination of dose rates from these isotopes, detector response curves, the thiecness, of the MS line, and the geometry of the MS line relative to the detector, the dose rate equivalent to MS line ooncentration is obtained.

The quantity of radi6active effluents released is obtained by multiplying the xenoan 133 equivalent MS line concentrations by the isotope ratio times the steam release rate, These detectors are s'aety-related Class 1E and meet the requirements of RG 1.97 and NUREG-0737, nlne monitors and are adjacent.

to the linves jien tIoietlii -dhCuid(IVCl.fhM).

.. -'he monitors are used as an aid In determining the' source of SO blowdown radioactivity due to SO tube mpture or a large' primaryto-secondary leak.T-hes~e-de~teors-are-sfety-relatedCass4E-and-flet-t~he-requhemtffts-Of-PRG--9-7..

' z4 11.,52.5,5 Main Steam Lne High Bnergy. Gamm-a (N-16) Monitors:

'Each main steatn line is monitored by an ATL Nat scintillatiori detector.

These detectors were installed to monitor the status of steam generator primary' to secondary tube leaks and to provide a diagnostic tool for all individuals concerned with steam generator condition.

These detectors are designed to detect high energy gamma ictlvlit in the 6 to 7,2 MEV energy range. High energy gamma activity in the main k( ' steam lines indicates the presence of N-16. The level of N-16 in the main steam lines is used to 11.5-14 Revision 14 RS2 STPEGS UFSAR The new fuel assemblies are transported to the new fuel storage pit or to the new fuel elevator by the 15/2-ton, dual-service FHB crane. The 2-ton hoist of this crane is designed to handle new fuel assemblies.

New fuel handling is discussed in detail in Section 9.1.4. Use of the 2-ton hoist of the 15/2-ton crane or of the fuel-handling machine to handle new fuel ensures that the design uplift of the racks will not be exceeded.The new fuel storage pit is situated in the approximate center of each FHB. The floor of the new fuel storage pit is at El. 50 ft-3 inches. The new fuel storage pit access hatch is provided with a three-section protective cover at El. 68 ft. The fuel assemblies are loaded into the new fuel storage racks through the top and stored vertically.

9.1.1.3 Safety Evaluation.

Units 1 and 2 of the STPEGS are each provided with separate and independent fuel handling facilities.

Flood protection of each FHB is discussed in Section 3.4.1. Flooding of the new fuel storage pit from fluid sources inside either FHB is not considered credible since all fluid systems components are located well below the elevation of the new fuel storage pit access hatch. A floor drain is provided in the new fuel storage pit to minimize collection of water.The applicable design codes and the ability of the FHB to withstand various external loads and forces are discussed in Section 3.8.4. Details of the seismic design and testing are presented in Section 3.7.Missile protection of the FHBs is discussed in Section 3.5. Failure of nonseismic systems or structures will not decrease the degree of subcriticality provided in the new fuel storage pit.In accordance with American National Standards Institute (ANSI) N 18.2, the design of the normally dry new fuel storage racks is such that the effective multiplication factor will not exceed 0.98 with fuel of the highest anticipated enrichment in place, assuming optimum moderation (under dry or fogged conditions).

For the unborated flooded condition, assuming new fuel of the highest anticipated enrichment in place, the effective multiplication factor does not exceed 0.95. Credit may be taken for the inherent neutron-absorbing effect of the materials of construction.

The new fuel assemblies are stored dry, the 21-in. spacing ensuring a safe geometric array. Under these conditions, a criticality accident during refueling and storage is not considered credible.Consideration of criticality safety analysis is discussed in Section 4.3.Design of the facility in accordance with RG 1.13 ensures adequate safety under both normal and postulated accident conditions.

The new fuel storage racks also meet the requirements of General Design Criterion (GDC) 62.9.1.2 Spent Fuel Storage 9.1.2.1 Design Bases. The spent fuel pool (SFP) is a stainless steel-lined reinforced concrete pool and is an integral part of each FHB. All spent fuel racks are classified as seismic Category 1, as defined by RG 1.29, and as ANS SC 3.The spent fuel storage facility provides storage capacity for 1,969 high density absorber spent fuel racks in a honeycomb array in each unit. Two storage regions are provided in the SFP. Two of the 9.1-2 Revision 16 STPEGS UFSAR Region 2 rack modules on the southend of the pool (modules #12 and #16) have not been installed.

A Fuel Ultrasonic Cleaning system may be used in the open space designated for modules #12 and#16. The Fuel Ultrasonic Cleaning system is freestanding and is seismically qualified.

It has no adverse effect on the fuel assemblies that are selected for cleaning; nor does it have an adverse effect on the design function of the spent fuel pool or its associated support systems. Figure 9.1.2-2 shows the pool layout for both Units 1 and 2. The six Region I rack modules are located in the northwest comer of the spent fuel pool.The Region 1 racks have 10.95-in.

nominal center-to-center spacing between the cells. This region is conservatively designed to accommodate unirradiated fuel at enrichments to 4.95 weight percent.Region 1 storage cells are each bounded on four sides by a water box except oil the periphery of the pool. The Region 1 spent fuel racks include a lead-in-guide to assist in depositing fuel assemblies into the fuel cell. Figure 9.1.2-3 shows a typical Region 1 spent fuel rack.The reactivity characteristics of fuel assemblies which are to be placed in the spent fuel storage racks are determined and the assemblies are categorized by reactivity.

Alternately, if necessary, all assemblies may be treated as if each assembly is of the highest reactivity class until the actual assembly reactivity classification is determined.

Section 5.6 of the Technical Specifications provides the definitions of the reactivity classifications and the allowed storage patterns.

Fuel assemblies are loaded into theracks in a geometrically safe configuration to ensure rack subcriticality.

Fuel assembly reactivity requirements for close packed storage and checkerboard storage are specified in the Technical Specifications.

The boron concentration of the water in the spent fuel pool is maintained at or above the minimum value needed to ensure that the rack Keff is less than or equal to 0.95 in the event of misplaced assemblies in the close packed storage .areas or in checkerboard storage areas. Consideration of criticality safety is discussed in Section 4.3.The Region 2 racks have a 9.15-in. nominal center-to-center spacing with fixed absorber material surrounding each cell. A sheet of neutron absorber material is captured between the side walls of all adjacent boxes. To provide space for the absorber sheet between boxes, a double row of matching flat round raised areas are coined into the side walls of all boxes. The raised dimension of these locally formed areas on each box wall is half the thickness of the absorber sheet. The boxes are fusion welded together at all these local areas. The absorber sheets are scalloped along their edges to clear these areas. Figure 9.1.2-4 shows a typical Region 2 spent fuel rack.The axial location of the absorber with respect to the active fuel region is provided and maintained by the structure of each box. At the outside periphery of each rack, a sheet of absorber material is captured under thin stainless sheets which are intermittently welded all around to the box.All rack surfaces that come into contact with fuel assemblies are made of annealed austenitic stainless steel. These materials are resistant to corrosion during normal and emergency water quality conditions.

The racks are designed to withstand normal operating loads as well as to remain functional with the occurrence of an SSE. The racks are designed with adequate energy absorption capabilities to withstand the impact of a dropped spent fuel assembly from the maximum lift height of the spent fuel pit bridge hoist. The racks are designed to withstand a maximum uplift force equal to the uplift force of the bridge hoist. The 14-in. and 16-in. racks also meet the requirements of ASME Code,Section III, Appendix XVII. The high-density spent fuel racks meet the criteria of Appendix D to Standard Review Plan (SRP) 3.8.4.9.1-3 Revision 16 STPEGS UFSAR Shielding for the SFP is adequate to protect plant personnel from exposure to radiation in excess of published guideline values as stated in Section 12.1. A minimum depth of approximately 13 ft of water over the top of an array of 193 (full core) assemblies with 42 hours4.861111e-4 days <br />0.0117 hours <br />6.944444e-5 weeks <br />1.5981e-5 months <br /> of decay is required to limit radiation from the assemblies to 2.5 mR/hr. or less.The FHB Ventilation Exhaust System is designed to limit the offsite dose in the event of a significant release of radioactivity from the fuel, as discussed in Sections 12.3.3, 15.7.4, and 9.4.2. However, no credit for the FHB Ventilation Exhaust System is taken in the LOCA and Fuel Handling accident in Chapter 15.The FHB is designed to prevent missiles from contacting the fuel. A more detailed discussion on missile protection is given in Section 3.5.In addition, space is provided for storage of fuel during refueling inside the RCB for 64 fuel assemblies in four 4 x 4 modules having 16-in, center-to-center spacing (Figure 9.1.2-1A).

These modules are firmly bolted in the floor.9.1.2.2 Facilities Description.

The FHB abuts the south side of the RCB and is adjacent to the west side of the MEAB of each unit. The locations of the two FHBs are shown in the station plot plan on Figure 1.2-3. For general arrangement drawings of the spent fuel storage facilities, refer to Figures 1.2-39 through 1.2-48 as listed in Table 1.2-1.The spent fuel storage facilities are designed for the underwater storage of spent fuel assemblies and control rods after their removal from the reactor vessel. The spent fuel is transferred to the FHB and handled and stored in the spent fuel pool underwater.

The fuel is stored to permit some decay, then transferred offsite. For a detailed discussion of spent fuel handling, see Section 9.1.4.The SFP is located in the northwest quadrant of each FHB. The floor of the pool is at El. 21 ft-1 I in., with normal water level at El. 66 ft-6 inches. The top of a fuel assembly in a storage rack does not extend above the top of the storage rack which is El. 39 ft-10 in. maximum. The fuel assemblies are loaded into the spent fuel racks through the top and are stored vertically.

9.1.2.3 Safety Evaluation.

Units 1 and 2 of the STPEGS are each provided with separate and independent fuel handling facilities.

Flood protection of each FHB is discussed in Section 3.4.1.A detailed discussion of missile protection is provided in Section 3.5.The applicable design codes and the various external loads and forces considered in the design of the FHB are discussed in Section 3.8.4. Details of the seismic design and testing are presented in Section 3.7..Design of this storage facility in accordance with GDC 62 and RG 1.1 3 ensures a safe condition under normal and postulated accident conditions.

The Keff of the spent fuel storage racks is maintained less than or equal to 1.00, even if unborated water is used to fill the spent fuel storage pool, by both the designs of the fuel assemblies and the storage rack and the use of administrative procedures to control the placement of burned and fresh fuel and control rods.Under accident conditions, the Kefr is maintained well below 0.95 assuming 2200 ppm borated water.The boron concentration of the water in the spent fuel pool is maintained at or above the minimum 9.1-4 Revision 16 REQUIREMENTS FOR RELIABLE SPENT FUEL POOL LEVEL INSTRUMENTATION AT OPERATING REACTOR SITES AND CONSTRUCTION PERMIT HOLDERS All licensees identified in Attachment 1 to this Order shall have a reliable indication of the water level in associated spent fuel storage pools capable of supporting identification of the following pool water level conditions by trained personnel: (I) level that is adequate to support operation of the normal fuel pool cooling system, (2) level that is adequate to provide substantial radiation shielding for a person standing on the spent fuel pool operating deck, and (3) level where fuel remains covered and actions to implement make-up water addition should no longer be deferred.1 .The spent fuel pool level instrumentation shall include the following design features: 1..,1 Instruments:

The instrumentation shall consist of a permanent, fixed primary instrument channel and a backup instrument channel. The backup instrument channel may be fixed or portable.

Portable instruments shall have capabilities that enhance the ability of trained personnel to monitor spent fuel pool water level under conditions that restrict direct personnel access to the pool, such as partial structural damage, high radiation levels, or heat and humidity from a boiling pool.1.2 Arrangement:

The spent fuel pool level instrument channels shall be arranged in a manner that provides reasonable protection of the level indication function.-* against missiles that may result from damage to the structure over the spent fuel pool. This protection may be provided by locating the primary instrument channel and fixed portions of the backup instrument channel, if applicable, to maintain instrument channel separation within the spent fuel pool area, and to utilize inherent shielding from missiles provided by existing recesses and corners, in the spent fuel. pool structure.

1.3 Mounting

Installed instrument channel equipment within the spent fuel pool shall be mounted to retain its design configuration during and following the maximum seismic ground motion considered in the design of the spent fuel pool structure.

1.4 Qualification

The primary and backup instrument channels shall be reliable at temperature, humidity, and radiation levels consistent with the spent fuel pool water at saturation conditions for an extended period. This reliability shall be established through use of an augmented quality assurance process (e.g., a process similar to that applied to the site fire _p6tection program).1.5 Independence:

The primary instrument channel shall be independent of the backup instrument channel.1.6 Power supplies:

Permanently installed instrumentation channels shall each be powered by a separate power supply. Permanently installed and portable instrumentation channels shall provide for power connections from sources independent of the plant ac and dc power distribution systems, such as portable generators or replaceable batteries.

Onsite generators used as an alternate power source and replaceable batteries used for instrument channel power shall have sufficient capacity to maintain the level indication function until offsite Attachment 2 resource availability is reasonably assured.1.7 Accuracy:

The instrument channels shall maintain their designed accuracy following a power interruption or change in power source without recalibration.

1.8 Testing

The instrument channel design shall provide for routine testing and calibration.

1.9 Display

Trained personnel shall be able to monitor the spent fuel pool water level from the control room, alternate shutdown panel, or other appropriate and accessible location.

The display shall provide on-demand or continuous indication of spent fuel pool water level.2. The spent fuel pool instrumentation shall be maintained available and reliable through appropriate development and implementation of the following programs: 2.1 Training:

Personnel shall be trained in the use and the provision of alternate power to the primary and backup instrument channels.2.2 Procedures:

Procedures shall be established and maintained for the testing, calibration, and use of the primary and backup spent fuel pool instrument channels.2.3 Testing and Calibration:

Processes shall be established and maintained for scheduling and implementing necessary testing and calibration of the primary and backup spent fuel pool level instrument channels to maintain the instrument channels at the design accuracy.

(NEI 12-02 (Revision 1)-.. Avust 2012 The three critical levels that must be monitored in a spent fuel pool are discussed below.It should be noted that continuous indication from a single instrument over the entire span from level 1 to level 3 is not required but could be used. If more than one instrument is used to monitor the entire span, that set of instruments constitutes a single channel satisfying either the primary or backup instrument channel requirement (refer to Figure 1 below).A visual representation of monitoring levels 1, 2 and 3 and the associated requirements for monitoring between the points are presented in Figure 1. The minimum requirements apply to the separation distance between level indications and support development of appropriate response procedures.

These requirements are separate firom the instrument channel design accuracy discussed in section 3, which apply to either discrete or to continuous instruments.

Figure 1 2.3.1. Level 1 -level that is adequate to support operation of the normal fuel pool cooling system A typical fuel pool cooling system design includes a combination of weirs and/or vacuum breakers that prevent siphoning of the pool water level, below a minimum level, in the event of a piping rupture that can affect the SFP level.Level 1 represents the HIGHER of the following two points: 3 NEI 12-02 (Revision 1)August 2012' The level at which reliable suction loss occurs due to uncovering of the coolant inlet pipe, weir or vacuum breaker (depending on the design), or The level at which the water height, assuming saturated conditions, above the centerline of the cooling pump suction provides the required net positive suction head specified by the pump manufacturer or engineering analysis.This level will vary from plant to plant and the instrument designer will need to consult plant-specific design information to determine the actual point that supports adequate cooling system performance.

2.3.2. Level

2 -level that is adequate to provide substantial radiation shielding for a person standing on the spent fuel pool operating deck Level 2 represents the range of water level where any necessary operations in the vicinity of the spent fuel pool can be completed without significant dose consequences from direct gamma radiation from the stored spent fuel. Level 2 is based on either of the following:

10 feet (+/- 1 foot) above the highest point of any fuel rack seated in the spent fuel pools, or a designated level that provides adequate radiation shielding to maintain personnel radiological dose levels within acceptable limits while performing local operations in the vicinity of the pool. This level shall be based on either plant-specific or appropriate generic shielding calculations, considering the emergency conditions that may apply at the time and the scope of necessary local operations, including installation of portable SFP instrument channel components.

Additional guidance can be found in EPA-400 (Reference 4), USNRC Regulatory Guide 1.13 (Reference

5) and ANSI/ANS-57.2-1983 (Reference 6).Designation of this level should not be interpreted to imply that actions to initiate water make-up must be delayed'until SFP water levels have reached or are lower than this point.2.3.3. Level 3 -level where fuel remains covered and actions to implement make-up water addition should no longer be deferred.Level 3 corresponds nominally (i.e., +/- 1 foot) to the highest point of any fuel rack seated in the spent fuel pool. Level 3 is defined in this manner to provide the maximum range of information to operators, decision makers and emergency response personnel.

Designation of this level should not be interpreted to imply that actions to initiate water make-up must or should be delayed until this level is reached.4 Nuclear Operating Company South 7=zr FPro/cct Bfectr/c G&neMA/ng Statlon 10P Box 28,9 W1 qdsworth, r.as 774,63 February 28, 2013 NOC-AE-13002959 10 CFR 50.4 10 CFR 2.202 U, S. Nuclear Regulatory Commission Attention:

Document Control Desk Washington, DC 20555-0001 South Texas Project Units 1&2 Docket Nos. STN 50-498, STN 50-499 Overall Integrated Plan Regarding Commission Order Modifying Licenses with Regard to Reliable Spent Fuel Pool Instrumentation (Order Number EA-12-051)

References:

1. Letter, Eric Leeds to E. D. Halpin, "Issuance of Order to Modify Licenses with Regard to Requirements for Reliable Spent Fuel Pool Instrumentation," March 12, 2012 (EA-12-051)

2. NRC Interim Staff Guidance JLD-ISG-2012-03, Compliance with Order EA-12-051, Reliable Spent Fuel Pool Instrumentation, Revision 0, August 29, 2012 3. Letter D. W. Rencurrel to NRC, "Initial Status Report in Response to March 12, 2012 Commission Order Modifying Licenses with Regard to Requirements for Reliable Spent Fuel Pool Instrumentation (Order Number EA-12-051)", dated October 24., 2012 On March 12, 2012, the Nuclear Regulatory.Commission (NRC) issued an order (Reference 1)to STP Nuclear Operating Company (STPNOC).

Reference 1 directs sTP Nuclear Operating Company to provide a reliable indication of the water level in associated spent fuel storage pools. Specific requirements are outlined in Attachment 2 of Reference 1.Reference 1 required submission of an overall integrated plan, including how compliance will be achieved.

The final interim staff guidance (Reference

2) was issued August 29, 2012 providing licensees an acceptable approach for complying with the order. The purpose of this letter is to provide the overall integrated, plan, Including a description of how compliance will be achieved pursuant to Section IV, Condition C.1.a, of Reference 1 in accordance with the guidance In Attachment 2 to Reference I and the guidance in Reference

2. See the Enclosure for STPNOC's response to the requested information.

There are no new commitments in this letter.33650640 NOC-AE-1 3002959 Page 2 of 3 If there are any questions regarding this letter, please contact Robyn Savage at (361) 972-7438.I declare under penalty of perjury that the foregoing is true and correct.Executed on: Ocflcl' el/, Dennis L. Koehl President and CEO/CNO

Enclosure:

South Texas Project (STP) Overall Integrated Plan for Implementation of Unit 1 &Unit 2 Spent Fuel Pool Level Instrumentation to Meet NRC Order EA-12-051 33650640 NOC-AE-13002959 Page 3 of 3!i cc: (paper copy)(electronic copy)Regional Administrator, Region IV U. S. Nuclear Regulatory Commission 1600 East Lamar Boulevard Arlington, TX 76011-4511 Balwant K. Singal Senior Project Manager U.S. Nuclear Regulatory Commission.

One White Flint North (MS 8 B1)11555 Rockville Pike Rockville, MD 20852 NRC Resident Inspector U. S. Nuclear Regulatory Commission P. 0: Box 289, Mail Code: MNI 16 Wadsworth,-TX 77483 C. M. Canady City of Austin Electric Utility Department 721 Barton Springs Road Austin, TX 78704 U. S. Nuclear Regulatory Commission Director of Office of Nuclear Regulation One White Flint North (MS 13 H 16M)11555 Rockville Pike Rockville, MD 20852-2738 A. H. Gutterman, Esquire Morgan,. Lewis & Bockius LLP Balwant K. Singal U. S. Nuclear Regulatory Commission

• John Ragan Chris O'Hara Jim von Suskil NRG South Texas LP Kevin Polio Richard Pena City Public Service Peter Nemeth Crain Caton & James, P.C.C. Mele City of Austin Richard A. Ratliff Texas Department of State Health Services Alice Rogers Texas Department of State Health Services Q.9 33650640 ENCLOSURE NOC-AE-13002959 South Texas Project (STP)Overall Integrated Plan for Implementation of Unit 1 & Unit 2 Spent Fuel Pool Level Instrumentation to Meet NRC Order EA-12-051 (S Page 1 of 12 STP Overall Integrated Plan for Implementation of Unit 1 & Unit 2 Spent Fuel Pool Instrumentation to Meet NRC Order EA 12-051 Revision:

00 1.0 OVERALL INTEGRATED PLAN INTRODUCTION This document provides the overall Integrated Plan (the "Plan") which the STP Nuclear Operating Company ("STPNOC")

will implement for Units I and 2 to comply with the requirements of NRC Order EA-12-051, "Issuance of Order to Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation" (Ref.2), (the "ORDER"), NRC Interim Staff Guidance JLD-ISG-2012-003

[Rev.0] (Ref.3), (the "ISG"), and NEI Report 12-02[Rev.1] ("NEI 12-02").This Plan follows the format and provides all of the information on the STP 1 & 2 Integrated Plan that is required in NEI 12-02 [Rev.1] (Ref.1), Section A-2-2. Throughout this Plan, any reference to NEI 12-02 and the ISG will be based on the revisions above.Any reference to NEI 12-02 will include compliance to the clarifications and exceptions to NEI 12-02 required by the Interim Staff Guidance, Rev. 0.In response to the NRC requirements, STPNOC will provide two channels of independent, permanently-installed, wide-range spent fuel pool level instrumentation

("SFPLI"), for the spent fuel pool ("SFP") of each unit. The spent fuel pool for each unit is independent and not interconnected in any way. For each SFP, the instrumentation provided for each channel will utilize the same technology, as permitted by the NEI 12-02 [Rev.1]. The spent fuel pool level instrumentation will provide continuous level indication for each SFP on both the Primary and Backup Channels.Both the Primary and Backup Channel/Instrument location and display of the SFP level will be independently mounted in each-unit's Radwaste Control Room in the Mechanical Electrical Auxiliary Building (MEAB), which is an accessible post-event location.

Other locations are still being considered.

Both the Primary and Backup Channel remote, non-safety related indication of the SFP level will also be provided in each unit's Control Room via input to the Plant Computer.The instrumentation systems will not be safety-related, but will meet the requirements for augmented quality in accordance with NEI 12-02 [Rev. 1] and the ISG as described below.Since all of the potential suppliers have not completed development, the information in this Plan is based on the overall .strategy and on information which, based on current information from potential suppliers, is thought to envelope the systems being developed for this application.

If there are any changes to the requirements in NRC JLD-ISG-2012-003

[Rev.0] and NEI 12-02 [Rev.1], relief from the requirements and schedule documented in this Plan may be required, in accordance with Section 12.0. Any required changes to this Plan will be defined in the periodic status reports submitted to the NRC.2.0 APPLICABILITY:

This Plan applies to the spent fuel pools for South Texas Project Unit 1 and Unit 2.Page 2 of 12 STP Overall Integrated Plan for Implementation of Unit 1 & Unit 2 Spent Fuel Pool Instrumentation to Meet NRC Order EA 12-051 Revision:

00 3.0 SCHEDULE: The installation of reliable spent fuel pool level instrumentation for the spent fuel pool associated with Unit 1 is scheduled for completion prior to 10/28/2015, which is the end of the second refueling outage (1 REI 9) following submittal of this Plan.The installation of reliable spent fuel pool level instrumentation for the spent fuel pool associated with Unit 2 is scheduled for completion prior to 4/29/2015, which is the end of the second refueling outage (2RE17) following submittal of this Plan.Unit 1 Milestones are as follows:* Design/Engineering

-September of 2014* Purchase of instruments

& equipment

-February of 2015* Receipt of equipment

-June of 2015* Unit 1 Installation

& Functional Testing -October of 2015 Unit 2 Milestones are as follows:* Design/Engineering

-December of 2013 ( .. ", Purchase of instruments

& equipment

-August of 2014* Receipt of equipment

-November of .2014* Installation

& Functional Testing -April of 2015 Consistent with the requirements of the ORDER and the guidance from NEI 12-02 [Rev.1], status reports will be generated in six (6) month intervals from the submittal of this Plan.4.0 IDENTIFICATION OF SPENT FUEL POOL WATER LEVELS: The STP Unit I and 2 spent fuel pools are essentially identical.

The following SFP elevations are identified:

• The bottom of the pool is at Plant El. 21 ft. 11 in.* The top of the SFP racks is approximately at Plant El. 39 ft. 10 in.* The minimum Limiting Conditionfor Operation SFP level is Plant El. 62 ft.* Normal SFP water level is at Plant El. 66 ft. 6 in.* Non-safety related level switch alarms are activated at Plant El. 67 ft. on high level and Plant El. 66 ft. on low level.* The spent fuel pool deck is at Plant El. 68 ft.The required key SFP water levels, per guidance of NEI 12-02 [Rev.1] and ISG (with clarifications and exceptions), are as follows: Page 3 of 12 STP Overall Integrated Plan for Implementation of Unit 1 & Unit 2 Spent Fuel Pool Instrumentation to Meet NRC Order EA 12-051 Revision:

00 4.1 LEVEL 1: Level adequate to support operation of the normal fuel pool cooling system.LEVEL 1 represents the HIGHER of either the level at which reliable suction loss to the cooling pump occurs, or, the required NPSH (Nominal Pump Suction Head) of the cooling pump.*Loss of reliable suction to SFP cooling pumps. For the purposes of this Plan, this level will conservatively be placed at Plant El. 64 ft. 2 in. This. allows for just over 1 ft. of SFP water level above the top of the suction inlet flange (SFP Cooling Pump 14 in. suction line with centerline of suction inlet flange at Plant El. 62 ft. 6 in.)which will be sufficient for NPSH. (Ref. 9)Therefore, considering the top of SFP fuel storage rack is at Plant El. 39 ft. 10 in., the indicated level on either the Primary or Backup Instrument Channel of greater than 24 ft. 4 in, above the top of the SFP fuel storage racks based on the design accuracy for the instrument channel per NEI 12-02 [Rev.1], for both the Primary and Backup Instrument Channels, is adequate for normal SFP cooling system operation.

LEVEL 1 = Plant El. 64 ft. 2 in or 24 ft. 4 in. water level above the top of the SFP fuel storage rack 4.2 LEVEL 2: Level adequate to provide substantial radiation shielding for a person standing on the SFP operating deck.Indicated level on either the Primary or Backup Instrument Channel of greater than 10 ft, above the top of SFP stored fuel assemblies based on current guidance in NRC RG 1.13 [Rev.2] (Ref. 4) will achieve substantial radiation shielding.

Requirements on substantial SFP radiation shielding is also given in ANSI/ANS-57.2-1983 (Ref. 5), and states that radiation shall not exceed 2.5 mRem/hr, but the minimum water depth to achieve this is undefined.

NRC RG 1.13 [Rev.2] took exception to using dose rates as design input for minimum SFP water level, and instead defined the minimum level as 10 ft. above the stored fuel assemblies.

STPNOC elects to use the conservative approach of defining the top of the fuel rack as a basis for measurement.

Therefore, indicated level on either the Primary or Backup Instrument Channel of greater than 10 ft. above the top of the SFP fuel storage rack, based on the design accuracy of the instrument channel per NEI 12-02 [Rev.1], for both the Primary and Backup Instrument Channels, ensures there is adequate water level to provide substantial radiation shielding for a person standing on the spent fuel pool operating deck.Page 4 of 12 STP Overall Integrated Plan for Implementation of Unit I & Unit 2 Spent Fuel Pool Instrumentation to Meet NRC Order EA 12-051 Revision:

00 LEVEL 2 []Plant El 49 ft. 10 in. or 10 ft. water level above the top of the SFP fuel storage rack.4.3 LEVEL 3: Level where the fuel remains covered.As stated above, STPNOC elects to use the conservative approach of defining the top of the fuel rack as a basis for measurement.

The installation of the SFPLI sensor will be such that it will measure as close as possible to the top of the SFP fuel rack. Indicated level on either the Primary or Backup Instrument Channel of greater than 1 ft. above the top of SFP fuel storage racks based upon the design accuracy of the instrument channel per NEI 12-02 [Rev.1], for both the Primary and Backup Instrument Channels, satisfies the NEI 12-02 [Rev.2] requirement of +/- 1 ft.from the top of the fuel rack. This monitoring level ensures there is adequate water level above the stored fuel seated in the SFP fuel storage rack.LEVEL 3 = Plant El 40 ft. 4 in. or 6 in. water level above the top of the SFP fuel storage rack.5.0 INSTRUMENTS:

Both the Primary and Backup Instrument Channels will utilize permanently-installed instruments.

The design of the primary. and backup instruments will be consistent with the requirements by NEI 12-02 [Rev.1], the ISG, and this Plan.The current plan is for both channels to utilize Guided Wave Radar, which functions according to the principle of Time Domain Reflectometry (TDR). A generated pulse of electromagnetic energy travels down the probe. Upon reaching the liquid surface the pulse is reflected and based upon reflection times level is inferred.

The measured range will be continuous from the high pool level elevation (67') to the top of the spent fuel racks (Ref. 8). However, STP is still evaluating other designs for this application.

Any changes to the current plan will be reported in the 6 month update letter.The supplier for the SFP instrumentation will be chosen by a competitive bidding process completed after submittal of this Plan, so the information in this Plan is based on the overall strategy and on available information from potential supplier's information on systems being developed for this application.

5.1 Primary

(fixed) Instrument Channel The Primary Instrument Channel level sensing components will be located in the northeast corner of the Spent Fuel Pool, as shown in Attachment

1. The primary instrument channel will provide continuous level indication over a range from Plant El. 40 ft. 4 in. (LEVEL 3) to Plant El. 67 ft. (SFP high level alarm) or a range of 26 ft.8 in. In addition, the capability for discrete level indications at LEVEL1, LEVEL 2 and LEVEL 3, as described in Section 4.0, will be available.

Page 5 of 12 Nuclear Operating Company South 7Xas PlotlectrE/ic GeeiýIewiln Stjtioi RD , Box 2S.9 adsmoith, Tewos 77483 June 25, 2013 NOC-AE-1 3003008 File No.: G25 10 CFR 2.202 U. S. Nuclear Regulatory Commission Attention:

Document Control Desk Washington, DC 20555-0001 South Texas Project Unit 1 & 2 Docket Nos. STN 50-498, STN 50-499 Response to Request for Additional Information Regarding the Overall Integrated Plan in Response to Order EA-12-051,"Reliable.Spent Fuel Pool Instrumentation" (TAC Nos. MF0827 and MF0828)

References:

1. Letter, Eric Leeds to E. D. Halpin, "Issuance of Order to Modify Licenses with Regard to Requirements for Reliable Spent Fuel Pool Instrumentation," March 12, 2012 (EA-1 2-051) (ST-AE-NOC-1 2002271) (ML1 2054A679)2. Letter, D. L. Koehl to NRC Document Control Desk, "Overall Integrated Plan Regarding Commission Order Modifying Licenses with Regard to Reliable Spent Fuel Pool Instrumentation (Order Number EA-12-051)," dated February 28, 2013 (NOC-AE-1 3002959) (ML13070A006)

3. NRC letter dated June 7, 2013, "South Texas Project, Units 1 and 2 -Request for Additional Information RE: Overall Integrated Plan in Response to Order EA-12-051, "Reliable Spent Fuel Pool Instrumentation" (TAC Nos. MF0827 and MF0828) (ST-AE-NOC-13002439) (ML131149A09)

On March 12, 2012, the Nuclear Regulatory Commission (NRC) issued an Order (Reference 1)modifying licenses with regard to requirements for reliable spent fuel pool instrumentation.

On February 28, 2013, STP Nuclear Operating Company (STPNOC) submitted an Overall Integrated Plan (OIP) (Reference

2) in response to the NRC Order. By a letter (Reference 3)dated June 7,. 2013, the NRC staff determined that additional information is needed to complete their review of the OIP. The STPNOC response to Reference 3 is provided in the attachment to this letter.There are no regulatory commitments in this letter.STI: 33704694 I NOC-AE-13003008 Page 2 If there are any questions, please contact Ken Taplett at 361-972-8416.

1 declare under penalty of perjury that the foregoing is true and correct.Executed on: U'"-, 25 Z5oL G. T. Powell Site Vice President

Attachment:

Response, to Request for Additional Information Regarding Reliable Spent Fuel Pool Instrumentation (Order Number EA-12-051) kit NOC-AE-13003008 Page 3 cc: (paper copy)(electronic copy)Regional Administrator, Region IV U. S. Nuclear Regulatory Commission 1600 East Lamar Boulevard Arlington, TX 76011-4511 Balwant K. Singal Senior Project Manager U.S. Nuclear Regulatory Commission One White Flint North (MS 8 B1)11555 Rockville Pike Rockville, MD 20852 NRC Resident Inspector U. S. Nuclear Regulatory Commission P. 0. Box 289, Mail Code: MN1 16 Wadsworth, TX 77483 C. M. Canady City of Austin Electric Utility Department 721 Barton Springs Road Austin, TX 78704 U.SI Nuclear-Regula-tory ciission Director, Office of Nuclear Reactor Regulation One White Flint North (MS 13 H 16M)11555. Rockville Pike Rockville, MD 20852-2738 A. H. Gutterman, Esquire Morgan, Lewis & Bockius LLP Balwant K. Singal U. S. Nuclear Regulatory Commission John Ragan Chris O'Hara Jim von Suskil NRG South Texas LP Kevin Polio Richard Pefia City Public Service Peter Nemeth Crain Caton & James, P.C.C. Mele City of Austin-RihrdA.R tliff .Texas Department of State Health Services Robert Free Texas Department of State Health Services Attachment NOC-AE-1 3003008 Page 1 of 23 Response to Request for Additional Information Regarding Reliable Spent Fuel Pool Instrumentation (Order Number EA-12-051)

References:

1. Letter, D. L. Koehl to NRC Document Control Desk, "Overall Integrated Plan Regarding Commission Order Modifying Licenses with Regard to Reliable Spent Fuel Pool Instrumentation (Order Number EA-12-051)," dated February 28, 2013 (NOC-AE-13002959) (ML13070A006)

2. Letter, Eric Leeds to E. D. Halpin, "Issuance of Order to Modify Licenses with Regard to Requirements for Reliable Spent Fuel Pool Instrumentation," March 12, 2012 (EA-12-051) (ST-AE-NOC-12002271) (ML12054A679)

3. NRC Japan Lessons-Learned Project Directorate Interim Staff Guidance JLD-ISG-2012-03, "Compliance with Order EA-12-051, Reliable Spent Fuel Pool Instrumentation, Revision 0, August 29, 2012 (ML12221A339)

4. NEI 12-02, Industry Guidance for Compliance with NRC Order EA-12-051, "To Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation," Revision 1, dated August 2012 (ML1 22400399)Reference I provided the Overall Integrated Plan (OIP) which the STP Nuclear Operating Company ("STPNOC")

will implement for Units 1 and 2 to comply with the requirements of NRC Order EA-12-051, "Issuance of Order to Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation" (Reference 2), NRC Interim Staff Guidance JLD-ISG-2012-003, Revision 0, (Reference

3) and NEI Report 12-02, Revision 1 (Reference 4).As discussed in Reference 1, any changes to the requirements in NRC JLD-ISG-2012-003 or NEI 12-02 may require relief from the requirements and schedule documented in the OIP.As provided in the OIP, the Milestones for completing the design and engineering work for Unit I are September 2014 and for Unit 2 is December 2013.The following responses to the request for additional information are based on information developed to date. Any changes to the following information that occur after completing and approving the final design for reliable spent fuel pool instrumentation will be provided in the periodic 6-month status reports submitted to the NRC required by Order EA-12-051.

Attachment NOC-AE-1 3003008 Page 2 of 23 REQUEST FOR ADDITIONAL INFORMATION OVERALL INTEGRATED PLAN IN RESPONSE TO ORDER EA-12-051, "RELIABLE SPENT FUEL POOL INSTRUMENTATION" STP NUCLEAR OPERATING COMPANY SOUTH TEXAS PROJECT, UNITS 1 AND 2 DOCKET NOS. 50-498 AND 50-499 1.0 Introduction By letter dated February 28, 2013 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML1 3070A006), STP Nuclear Operating Company (STPNOC, the licensee), submitted an Overall Integrated Plan (OIP) in response to the March 12, 2012, U.S. Nuclear Regulatory Commission (NRC), Commission Order modifying licenses with regard to requirements for Reliable Spent Fuel Pool (SFP) Instrumentation (Order Number EA-1 2-051;ADAMS Accession No. ML12054A679) for South Texas Project (STP), Units 1 and 2. The NRC staff endorsed Nuclear Energy Institute (NEI) 12-02, "Industry Guidance for Compliance with NRC Order EA-12-051, to Modify Licenses with Regard to Reliable SFP Instrumentation,"*Revision 1 dated August-2012 (ADAMS -Accession No.-ML12240A307), with-exceptions as documented in Interim Staff Guidance (ISG) 2012-03, "Compliance with Order EA-12-051, Reliable SFP Instrumentation," Revision 0, dated August 29, 2012 (ADAMS Accession No. ML12221A339).

The NRC staff has reviewed the February 28, 2013, response by the licensee and determined that the following request for additional information (RAI) is needed to complete its technical review. Please provide the response to the following RAIs.

Attachment NOC-AE-13003008 Page 3 of 23 2.0 Levels of Required Monitoring The OIP states, in part, that LEVEL 1: Level adequate to support operation of the normal fuel pool cooling system.Plant El. 64 ft. 2 in or 24 ft. 4 in. water level above the top of the SFP fuel storage rack.LEVEL 2: Level adequate to provide substantial radiation shielding for a person standing on the SFP operating deck.Plant El 49 ft. 10 in. or 10 ft. water level above the top of the SFP fuel storage rack.LEVEL 3: Level where the fuel remains covered.Plant El 40 ft. 4 in. or 6 in. water level above the top of the SFP fuel storage rack....The installation of the SFPLI [spent fuel pool level instrumentation]

sensor will be such that it will measure as close as possible to the top of the SFP fuel rack.Indicated level on either the Primary or Backup Instrument Channel of greater than 1/2 ft. above the top of SFP fuel storage racks based upon the design accuracy of the instrument channel per NEI 12-02 [Rev,1], for both the Primary and Backup Instrument Channels, satisfies the NEI 12-02 [Rev.2] requirement of+1 ft. from the top of the fuel rack. This monitoring level ensures there is adequate water level above the stored fuel seated in the SFP fuel storage rack.NRC RAI-la Please provide the following:

a) For Level 1, please specify how the identified location represents the HIGHER of the two points described in the NEI 12-02 guidance for this level.STPNOC Response LEVEL 1 represents the HIGHER of either the level at which reliable suction loss to the spent fuel pool (SFP) cooling pump occurs, or the required net positive suction head (NPSH) of the SFP cooling pump Required NPSH.The SFP cooling pumps were analyzed for the conservative worst case operation of the SFP cooling pumps. Maximum values for line resistance, fluid temperature, suction flow Attachment NOC-AE-13003008 Page 4 of 23 and static head were used to calculate NPSH parameters for both required and available NPSH (NPSHR and NPSHA). It was determined that for the worst case scenario, the NPSHA was significantly higher than NPSHR. The NPSHA was calculated to be 42.67 feet (ft) and NPSHR was calculated to be 18.75 ft.Therefore, NPSHR is not the determining value to be used for LEVEL 1.Loss of reliable suction to SFP cooling pumps.For the purposes of the OIP, this level is conservatively placed at Plant elevation (El.). 64 ft, 2 inches (in). This level provides for more than one foot of water above the top of the SFP cooling pump suction inlet flange (the centerline of the 14 inch suction line flange to the pump is at Plant El. 62 ft. 6 in.) which will be sufficient for NPSH.A vortex calculation shows 0.134% air entrainment at an elevation one foot above the suction pipe centerline.

Level 1 at 64 ft. 2 in. is adequate for normal SFP cooling system operation.

Therefore, Level 1 represents the HIGHER of the two points described in the NEI 12-02 guidance.NRC RAI-Ib b) A clearly labeled sketch depicting the elevation view of the proposed typical mounting arrangement for the portions of instrument channel consisting of permanent measurement channel equipment (e.g., fixed level sensors and/or stilling wells, and mounting brackets).

Please indicate on this sketch the datum values representing Level 1, Level 2, and Level 3 as well as the top of the fuel. Indicate on this sketch the portion of the level sensor measurement range that is sensitive to-measurement of the fuel pool level, with respect to the Level 1, Level 2, and Level 3 datum points.STPNOC Response See Figures 1 and 2 of this Attachment.

3.0 Instrumentation

and Design Features 3.1 Instruments and Arrangement The OIP states, in part, that Both the Primary and Backup Instrument Channels will utilize permanently-installed instruments....

The Primary Instrument Channel level sensing components will be located in the northeast corner of the Spent Fuel Pool, as shown in Attachment 1 ....

Attachment NOC-AE-1 3003008 Page 5 of 23 The Backup Instrument Channel level sensing components will be located in the northwest corner of the Spent Fuel Pool, as shown in Attachment 1 ....The current Plan is to mount the supporting electronic instruments outside of the spent fuel pool area, to provide a more benign radiation and environmental conditions, and also provide for reasonable and accessible locations for operators.

SFP Primary and Backup Channel Level Instruments are currently planned to be located in Radwaste Control Room of the Mechanical Auxiliary Building (MAB);however, STPNOC is still evaluating other possible locations (i.e. relay room).NRC RAI-2 Please provide a clearly labeled sketch or marked-up plant drawing of the plan view of the SFP area, depicting the SFP inside dimensions, the planned locations/

placement of the primary and back-up SFP level sensor, and the proposed routing of the cables that will extend from the sensors toward the location of the read-out/display device.STPNOC Response See Figure 3 of this Attachment.

3.2 Mounting

The OIP states, in part, that Consideration will be given to the maximum seismic ground motion that occurs at the installation location for the permanently installed equipment which is documented in the UFSAR [Updated Final Safety Analysis Report] Section 3.7.The mountings shall be designed consistent with the highest safety or seismic classification of the SFP. The level sensors will be mounted on seismically qualified brackets.NRC RAI-3a Please provide the following:

a) The design criteria that will be used to estimate the total loading on the mounting device(s), including static weight loads and dynamic loads. Please describe the methodology that will be used to estimate the total loading, inclusive of design basis maximum seismic loads and the hydrodynamic loads that could result from pool sloshing or other effects that could accompany such seismic forces.

m CD 0 CD 0 CD S-n 0-. ~0 0 0 R.CD CD CD z 0 CD)CD -CD\CD Attachment NOC-AE-13003008 Page 21 of 23 Figure 2 Proposed Mounting Arrangement A./1 RADAR HORN ANTENNA 4/AI VIEW A-A RG1 CALC. NO.STPNOC013-CALC-002 E N E R C 0 N CALCULATION COVER SHEET EvalAncG-Evefy Nr9. 1f t49 PAGE NO. I of 49 Title: Radiological Release Thresholds for Emergency Action Levels Client: South Texas Project Project: STPNOC013 Item Cover Sheet Items I Does this calculation contain any open assumptions that require confirmation? (If YES, Identify the assumptions) 2 Does this calculation serve as an "Alternate Calculation"? (If YES, Identify the design verified Calculation.)

Design Verified Calculation No.3 Does this calculation Supersede an existing Calculation? (If YES, identify the superseded Calculation.)

Superseded Calculation No.Scope of Revision:

Incorporate decay time of one hour from shutdown as well as migration into Attachment I. Change statement of no decay in the STAMPEDE runs,_____________

I _____________________________________

--Revision Impact on Results: Values calculated in Attachment 1 decreased and have become the limiting values.m"" m u~~~~~ &A a1 -- i-le~- lfin kZ ------I I L__Safety-Related E Non-Safety Related (Print Name and Sign)Originator:

Caleb Trainor Date: 3/21/2014 Design Verifier:

Chad Cramer Date: 3/21/14 Approver:

Date: 3J21/14 Marvin Morris CALC. NO. STPNOC013-CALC-002 ENERCON CALCULATION

&cd. , project Everyy. REVISION STATUS SHEET REV. 1 PAGE NO. 2.of 49 CALCULATION REVISION STATUS REVISION DATE, ...DESCRIPTION

• 0 03/03/20 14 Initial Issue 1 .3/21/2014 Resolve inconsistency in decay times for the two calculations PAGE, REVISION STATUS'PAGE NO. REVISION PAGE.NO. REVISION ATTACHMENT REVISiON STATUS r ATTACHMENT NO.PAGE NO. REVISION NO. ATTACHMENT NO. PAGE NO.REVISION NO.1 2 12-24 25-3 1.* .i .:1 ' .'. .1 ' .. = .., : -1.* i'3 32-40?

0E N ER C0N ExceIIefl,-EVVY toec.r NVy dty CALCULATION DESIGN VERIFICATION CHECKLIST CALC. NO. STPNOCOI3-CALC-002 REV. I PAGE NO. 3 of 49 CHECKLIST ITEMS Yes No N/A Design Inputs -Were the design inputs correctly selected, referenced (latest revision), consistent with the design basis and incorporated in the calculation?

Assumptions

-Were theassumptions reasonable and adequately described, ,/justified mad/or verified, and documented?

Quality Assurance

-Were the appropriate QA classification and requirements V/assigned to the calculation?

Codes, Standard and Regulatory Requirements

-Were the applicable codes, standards and regulatory requirements, including issue and addenda, properly V identified and their requirements satisfied?

Construction, and Operating Experience

-Have applicable construction and /operating experience been considered?

Interfaces

-Have the design interface requirements been satisfied, including interactions with other calculations?

Methods -Was the calculation methodology appropriate and properly applied to 1 _-sat-isfy-the.-calculation-objective?-

.............

Design Outputs -Was the conclusion of the calculation clearly stated, did it correspond directly with the objectives and are the results reasonable compared to the inputs?Radiation Exposure -Has the calculation properly considered radiation exposure to the public and plant personnel?

Acceptance Criteria -Are the acceptance criteria incorporated in the calculation sufficient to allow verification that the design requirements have been satisfactorily accomplished?

Computer Software -Is a computer program or software used, and if so, are the requirements of CSP 3.02 met?COMMENTS: None_____________________

I I: -.Print ,VameandSign)

Design Verifier:

Chad Cramer Date: 3/21/14 Others: Date:

'CALCULATION CALC. NO. STPNOC013-CALC-002 0 E NE R C O N DESIGN VERIFICATION REV. 1 PLAN AND

SUMMARY

SHEET PAGE NO. 4 of 49 Calculation Design Verification Plan: Calculation shall be verified by comparing the documented input with the references and checking the validity of the references for the intended use. As necessary, assumptionis shall be evaluated in'd verified to determine if they are based on sound engineering principles and practices.

Verify eapjiiable methodology, inputs, resuls, and conclusions, (Print Name and Sign for Approval -markc "NIA".if not required).l Apioe- Dat 3/21114..Calculation Design Verification Summafry:

___ __ __________

Design inputs, assumptions, methodology, res.... and conclusions were evaluatedlverified and found to be acceptable.

All comments have been incorporated.

Based On The Above Summary, The Calculation Is Determined To Be Acceptable.

Cii (Print Name and Sign)Design Verifier:

Chad Cramer Date: 3/21/14 Others: Date: C CALC. NO, STPNOCOI3-CALC-002 E N ERCO N ý Radiological Release Thresholds Pv.I Exceli-c#

,&,y proe Evey,&zy.

for Emergency Action Levels PAGE NO. 5 of 49 Table of Contents 1.0 OBJECTIVE/SCOPE

......................................................................................................................

6 2.0

SUMMARY

OF RESULTS ..................................................................................................................

6 3.0 METHOD OF ANALYSIS ....................................................................................................................

7 4.0 IN P U T S ...............................................................................................................................................

7

5.0 REFERENCES

......................................................................

.........................

7 6.0 ASSUMPTIONS

.......................................................

.........................

....................

8 7.0 STAMPEDE CALCULATIONS

....................................................................................................

9 7.1 U nusual E vent -R U I ...................................................................................................................................

9 7.2 Alert, Site Area and General Emergencies

-RA 1, RS1, RG1 ..............................................................

10 Attachment I -Hand Calculations

..................................................................................................................

12 Attachment 2 -Calculations'.........

..............

....................................

25 Attachment 3 -STAMPEDE OUTPUT ...........................................................................................................

32 CALC. NO. STPNOC013-CALC-002-E N E R C-:0 N Radiological Release Thre .sholds -_______I__

Smergency A evels PAGE NO. 6 of 49 1.0 OBJECTIVE/SCOPE The purpose of this calculation is to determine the Emergency Action Level (EAL) threshold values of a radiological release fiom the Unit Vent or Main Ste'm Lines for an Unusual Event, Alert, Site AteiEmefge1n cy,: 6o" General EmergeiXcy.

Th e calnulated threshold values are to be included in the STP EAL Technical Basis document, which implements the new NE 99.01, Revision 6, Emergency Action LeveliScheme and Will be sub iitted to tie NRC for approval' Uon NRC approval, the values will be used in OERPoI-ZV-IN0i,Revision l, .Ehmergency CLassification.

Both a hand calculation and the South Texas Assessment Model Projeotihg Emergency Dose Evaluation (STAMPEDt) sowaie prograinx¢re usedto geni'ate the results. The-hand calculation is included as Attachment 1.: Revision I of this calculation incorporated decay for a release taidng place one hour after reactor shutdown.

This was done to cteate continuity between the two methodologies iisent. : s u d w. Thi W:.'.' '"y. 0 s ..... .. ..- ..... ... ..;e.n-2.0

SUMMARY

OF RESULTS The results of the calculatiofis for the radiation monitors specified in the STP EAL Basis Document and are listed in Table 2.1, below. -_" ._. " ' '": "___....Table 2.1: Summary of Calculation Results,.C Emergency Action* Level s RT-80103B, Unit Vent (ACiCsec)RT-8046 through 8049, Main Steam Lines (tCi/cc)RU1t[ .Unusual Event ...(Alert*RG1 General Emergenicy

• STAMPEDE was not used to determine the threshold for RU'methodology should be used to determine the threshold Value.Reference 5.10 indicates that the ODCM This calculation will be used to establish the threshold values for abnormal radiation based emergencies in the STP EAL Technical Basis document..

(

CALC. NO. STPNOCO13-CALC-002

() EN ERCO N Radiological Release Thresholds RV, 1 xcelce-Ey project y do for Emergency Action Levels b~celeca-eiyrojeL fe~ydcyPAGE NO. 7 of 49 3.0 METHOD OF ANALYSIS Previously, STAMPEDE was used to calculate the Emergency Action Level threshold values for effluent releases.

A hand calculation will verify the STAMPEDE calculations.

The hand calculation is described in Attachment 1 of this document STAMPEDE conforms to the requirements of STP Procedure OPGP07-ZA-0014, Software Quality Assurance Program. STAMPEDE was run at STP on an STP computer and under the supervision of an ENERCON employee with access to the STP site as a critical worker.4.0 INPUTS 4.1 Per NEI 99-01, Revision 6, Initiating condition AU1, EAL 1, the Notice of Unusual Event initiating condition is a release of gaseous or liquid radioactivity greater than two times the ODCM limit for sixty minutes or longer (Reference 5.10).4.2 The ODCM offsite dose limit is exceeded if the Xe-133 release concentration exceeds 7.41E-04ýtCi/cc (Reference 5.6).4.3 The Unit Vent flow rate is 9.4E+07 cc/sec (Reference 5.1).4.4 The main steam line pressure and PORV choke flow rate are 1285 psig and 1.05E+06 lbmn/hr, respectively (Reference 5.2).-4-.5-The-specific-volume-ofosaturated-steanrat--1=285-pstg--0.3-38 ft-/lb-m-f(Rf--cT5.3).-

.- ..-4.6 The fglease concentration is varied to find the release concentration which correlates to each emergency action level. Emergency action levels are taken from NEI 99-01, Revision 6 (Reference 5.10) for initiating conditions AA1, ASI and AGI. EAL I is the EAL of interest in each initiating condition..

The doses at the Site Boundary that correlate to the threshold concentrations are listed in Table 4.1.Table 4.1 EAL Offsite Dose Initiating Conditions Alert Site Area General Thyroid CDE 50 mrem em 5000 mrerm

5.0 REFERENCES

i_ 1Off.site ].0datJion-Ma~nual-.-Res~s.~on-l-7, Mar-ah-2i 5.2 Main Steam PORV Capacity Verification MC05591, Revision:

I 5.3 NIST Steam Tables, 2011 5.4 0ERP01-ZV-1N01, Emergency Classification Draft Revision 10 5.5 OERPO1-ZV-TP01, Offsite Dose Calculations, Revision 21 5.6 STP Calculation NC-9012, CRMS Rad Monitor Setpoints, Revision 7 5.7 STP Calculation NC-901 1, Revision 2 5.8 STAMPEDE Computer Program, Revision 7.0.3.3 5.9 STAMPEDE User's Manual 5.10 NET 99-01, Revision 6, Development of Emergency Action Levels for Non-Passive Reactors 5.11 OPGP07-ZA-0014 Quality Assurance Program 5.12 ITWMS Call Number 1000010987 Design Document, Revision 0 r... ," ..' .: ..

CALC. NO. STPNOCO13-CALC-002 Radiological Release Thresholds CA-v.., 0 EIN E R.Cu NE. I_ _ __ _.for Emergency Action Levels.LB1Y. NO. 8 of 49 6.0 ASSUMPTIONS 6.1 Unit Vent Noble Gas Monitor To be coisisteat with thie ODCM methodology, the unit vent release is assumed tW be eniirely Xe-133. The unit vent noble gas monitor is calibrated to Xe-133 Referend*e5.

1) thefore;ie the irionitor reading accuratelyý reflects the Xe-133'release magnift &." To be consistent with ODCM methodology, the main steam line release is assumed to be entirely Xe-133. The noble gas monitor is calibrated to Xe-133 (Reference 5.6).6.2 Release Duration Per Reference 5.10, Sections IC AAl, AS-, and AGi developer notes, the release should be assumed to last onb hour.6.3 Release follow'ing Reactor Shutdown The release initiates one hourie'r reactor shutdown.

While"L release initi~ating aftreactor shutdown is likely, signifinct decay of short lived nu'ciide" occurs dUrbig the m7igration time. A release at reactor shitdow'i would have a significantiy higher activity at the monitor location than arthe-Teta iriti : it&-id urtant.nfo.rtli-e-thr.eshatod_!ddtnot-be.-ctalceulate_.tat-shutddwn.- -as.. this-would creýa ia very high thresh4Id Which woul idibtnb. appropriate for releases Which occur shortly after shutdown.

One hour afte'r shutdown is sufficient time to dec@' short lived nuclides and treate.a 'onservativethreshold. 6.4 Source Term Per Reference.

51, ahY unit vent r$glease Witic.eed RCS..ctivity

'nd no core melt should be calculated using tliegap iventory..

Thl fre, the gap.inventor.

is used fdr all.unit vent releases.Per Reference 5..1, for a main steam line releae following a steam generator tube rupture it is appropriate to use an inventory of 'noble gases plus 0.2% iodiiie. A steam generator tube rupture is the only scenario which would create significant dffsite doses through a main steam line release.6.5 Default STAMPEDE Input Values-0-d "veeoper n oe co pýfannttrad-SLt Ad-A-tn-dl-ýu ,gest-usirig-the ODCM or the site's emergency dose assessment methodology.

STAMPEDE is used for emergency dse' assessment.

Per Referen'e5.1, when actual meteorology is not available, the default STAMPEDE values should be used. Had t e'ODCM methodology been. used, the 500 hour0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br /> peak ,/Q value would be used whi'ch is less conservative than the y/Q value prodilced by STAMPEDE using default meteorological cbnditioni Therefore, the, use of STAMPEDE default values provides a more conservative estimate than that of the alterniative method outlined in Reference 5.10.6.6 Average Effluent Concentration (X/Q).The same X /Q is used for the unit vent and main steam line release. Reference

5.1 applies

the same unit vent XQ to Units I and 2 which would also be applicable to the main steam line. All releases are considered to be ground level releases.

CAL.C. NO. STPNOCO I 3-CALC-002 E N E A C N Radiological Release Thresholds for Emergency Action Levels REV. I~mce- Every project, Every day PAGE NO. 9 of 49 7.0 STAMPEDE CALCULATIONS

7.1 Unusual

Event -RU1 7.1.1 Unit Vent Monitor AUl recommends declaring an unusual event due to a release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or longer (Reference 5.10).STP sets the ODCM limit at 7.41E-04 jiCi/cc (Reference 5.6, pg. 16). Two times the limit would be 1.48E-03 ýtCi/cc..

The threshold is listed in jsCi/sec so that variations in flow rate do not change the threshold.

The normal flow rate from the unit vent is 9.4E+07 cc/sec (Reference 5.1).Concentration

• Flow Rate = Release Rate (1.48E- 03) * (9.4E + 07)scc) 1.4E + 05 Ise... -Eqzia2b-n 7.1.-71 -7.1.2 Main Steam Line Monitor The ODCM does not calculate a release corresponding to allowable limits for the main steam line monitors.

Since the unit vent release calculated in the ODCM was assumed to be primarily Xe-133, the assumption is made in the ODCM that other noble gases and iodine may be ignored in the calculation.

This assumption is equally justifiable for the main steam line and the same limiting release will be used.The magnitude of the release calculated for the unit vent Unusual Event applies to the main steam lines as well. The main steam line PORV's will create a dose exceeding two times the ODCM limit by releasing 1.4E+05 ýLCi/sec of activity which is equivalent to the release from the unit vent.The- steam lines hold saturated steam at 1285 psig, per Reference 5.2, which has a specific volume of 0.338 ft 3/lbm (Reference 5.3). The PORVs will release the steamn at 1.05E+06 lbm/hr per Reference 5.2. This creates a set flow rate of steam from the main steam lines of 2.79E+06 cc/sec as shown below.Flm (ftc 3 c F *Density bm -283168 ) sec--0/Tbm) (ft 3 fN /scecsc\c 1.05E + 06 *0.338 28316.846

+ 3600 (-e-) 2.79E +06-/Euaton7.12hr sec Equation 7.1.2.1 CALC. NO. STPNOC013-CALC-002 EN. E. R CO~NI Radiological Release Thresholds REV.I-." for Emergency Action Levels ErcwIC &PAGE NO. 10 of 49 C, Since the flow rate is set, the concentration will determine the limit. Equation 7.1.1.1 solves for the limiting concentration of 5.OOE-02 ýXCi/cc as shown below.Limi~ting Release ( Ci C Limtig elese(~ -Limiting Concentratiion (LI Release Rate (C) .scc)2 .00E -12 2.7 9

• 10 6 .CC( ', ." '. " :':. ... .': " ..'", " ' : :' : .." .Z "'.F .*.. :".: " : ..e Equation 7.1.2.2 7.2 Aler`t, Site Area and'General Emergencies RA1, RSi, RGI 7.2.1 Unit Vent Monitor Input'The Alert EAL is set to 16 inrem TEDE and 50 inrem Thyroid CDE per Reference 5.10.-.._. The~emegemcy~oiffsitost

.c ulationrsof war_ STAMPEDE was _usedQo calculate thle--release which corresponds to this dose. A release concenitration corr6lating to the EAL threshold valu u was calculated' by varying the input. The fhll1wingissiimptions and inputs were tised for'the calculation as described in'Sections 4,.0 and 6.0.o Releasebegins at redetor t"ip..Release. asts for one hour Gap..

rm.." Default STANTEDE input values O Windspeed 13.2 .-o:. Stability class D Results Given a monitored unit vent.release of 2.50E+06 0Ci/sec, the Thyroid CDE is 51 EFATInitiating.Cnonditinn ks exceeded.Threshold, values for the. Site& Aea Emergency ahd General E-nerkenctns are multiples of 10 and 100 of the Alert. Since'the correlatioh between release coceiintration and dose is linear, threshold values for the steam line monitors are 2.50E+07 and 2.50E+08 ýtCi/sec for the SAE and GE respectively.

Both are also limited by Thyroid CDE. Additional STAMPEDE iteratiols.

were performed to confirm this and are attached.The input and output files can be found at the end of this document in Attachment 3.(

CALC. NO. STPNOC013-CALC-002

...) E N E R C O N Radiological Release Thresholds RV.I for Emergency Action Levels PGN. I r'J c4~e~ijct~~~ePAGE NO. 11lof 49 7.2.2 Main Steam Line Monitor Input A release concentration correlating to the EAL threshold value was calculated by varying the input. The following assumptions and inputs were used for this calculation as described in Sections 4.0 and 6.0.o Release begins at reactor trip* Release lasts for one hour* Noble gas + iodine with 0.2% iodine source term* Default STAMPEDE input values o Windspeed

= 13.2 mph o Stability class D Results Given a monitored main steam line release of 4.5 ýtCi/cc, the Thyroid CDE is 50 mrem/hr and the EAL INrtiating Condition is exceeded..-...--..7... -The-input-and-output-fi~les-can be feund-at-the end.of-this-document in Attachment-3

.--7.3 Threshold values for the Site Area Emergency and General Emergencies are multiples of 10 and 100 of the Alert. Since the correlation between release concentration and dose is linear, threshold values for the steam line monitors are 45 and 450 ýLCi/cc for the SAE and GE respectively.

Both are also limited by Thyroid CDE. Additional STAMPEDE iterations were performed to confirm this and are attached.

Radiological Release Thresholds CALC. NO. STPNOC0.13-CALC-002 E N. EAR CO0N for Eriiergericy Action Levels .REV. I* -y P4-ecr..y " Attachment 1 PAGE NO. 12 of 49 Attachment 1 -Hand Calculations 1.0 OBJECTIVE/SCOPE Each release calýulated using STAMPEDE in the main. document is calculated by hand in this attachment and .he results compared to STAMPEDE.2.0

SUMMARY

OF RESULTS Table 2.1 is displayed again below showing th .results from all the calclilaiionhs The minor difference is due to STAMPEDE using decay factors*s overa"6ne-iripriod shutdown.

This also accounts for the change in the limiting dose being TEDE in the ha.id calculations' a Thyioid CDE in the STAMPEDB calculations.

The acduracy of. f the hand'calculation is c'hsidebed sutficient and recommended for use in Emer Action Leveis..Tablel2.1 Results Emergency Action .RT-8010b, Unit Vent. iRT8046 through 8049,....

-Main Steam-Line--

-RUt UnuisuAl Event C Ct II R(AI Aiert.* ' I.:.. ] :.'.I 14"bi I bite Area Emergency I .: 3.0 METHOD OF ANALYSIS Using the limiting dose at the site boundary, the release is back calculated using atmospheric dispersion models. The X/Q value used is calculated -from Regulatorj Guide 1.145, Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants. Rather than using the most conservative meteorology, average meteorological conditions are used as inputs C-Radiological Release Thresholds CALC. NO. S' PNOCOI3-CALC-002 E N E R C 0 N for Emergency Action Levels REV. I pro] Every do- Attachment 1 PAGE NO. 13 of 49 to most closely agree with STP emergency dose assessment methodology per the ODCM and STAMPEDE.

Assumed nuclide inventories are taken from Reference 5.4. The dose conversion factors aie taken from Reference 5.2. A release concentration is used to find an initial projected dose at the Site Boundary.

Using the projected dose at the Site Boundary, the release concentration is scaled to find the limiting dose for each EAL.4.0 INPUTS* The Unit Vent flow rate is taken from the Offsite Dose Calculation Manual; Revision 17, March 2011 and is 9.44E+07 cc/sec.* The main steam line pressure and PORV choke flow rate were taken from Reference 5.5 and are 1285 psig and 1.05E+06 Ibm/hr respectively.

• The specific volumLe of saturated steam at this pressure is taken from the NIST steam tables and is 0,338 ft 3/lbm.0 The release concentration is varied to find the release concentration which correlates to each emergency action level dose. Emergency action level doses are taken from NEI 99-01 Revision 6 for initiating conditions AA1, AS I and AGI. EAL I is the EAL of interest in each initiating condition.

The limiting doses are listed in Table 4.1. NEt 990-01 Revision 6 states that these valvalues ar e--b as edonractions-o otheEnvir ohent al-Prote tion=CggGi-e--Pt otecqiv-A-tio...

---Guidelines (EPA PAGs) and the General Emergency represents the protective action values recommended by the EPA.Table 4.1 EAL Thresholds Alert Site Area General Thyroid CDE _ 50 mrem.n 500 mrem O 5000 mrem o A release lasting one hour is selected per NEI 99-01 Revision 6 developer notes.* Atmospheric dispersion factors are calculated per Regulatory Guide 1.145 (Reference 5.1), The reactor building dimensions used as inputs for this calculation are taken from Reference 5.13.* Nuclide inventories are taken from TGX/THX 3-1, (Reference 5.4) which isthe source docunent for the nuclide inventories used in STAMPEDE.

The release inventories are a gap release and noble gases plus 0.2% iodine which are listed below. Each nuclide inventory was normalized to oto variCRI releaSe activities.

Radiological Release Thresholds CALC. NO. STPNOC013-CALC-002 E N E. R:C O N for Emeigency Action Levels .... 1L, Attachment 1 PAGE NO. 14 of 49 Table 4.2. Gap Inventory Ni...ide.

(6 6/cM Normaliked Activity Normalized" ' " z-3 1 SOtO 1 53E-0 Xe17 90--0 .9SO 1-132 .1.509E+/-05

.1.53E-03 .X-i 7.. .190E+61 1; .048-01 "" 1 .2.40E+05.05 2.45E-03 CS-134 ;70E+01 3.78E-07: Kr-S93n 1,.30E+0.6 1.33E,02 Te132.. .0. 49.E9108 Kr-85.. 3,70E+05 .8 3 Ru1 Q3, ,0 7 3 90OE1 K Zr9S 1.1OE-02 .... .12E4-l 0

......

.?:::~~ .. .0- .5 ..: ...fE-:0 ,.!:: .i:4.0 ..." Xe-7131m 10+0 12-0 CI4 I .74-01S03E Xe-133 2.20+07 2325E-01 Sr$.-, .6940E902.

6 55E910 C C T lbIe 4.3 Noble Gases+O.2%

loaine inventory Nuclide; Inventbry" -N6' i 'iz'd6d.....-132 .... 8:-.:.:.,..:

6 13-02 .;* 319.3-4:1134 1 06-0 6 2 X&31 280E-+00.;

.08,02',Xe-i3.3m 4,203.00.

.. 1 .. -.56E--02: .*-4) m-~-..--,---a..d-MAIfl4A4-.-

I A.tAJZflhl

't.UVJDUI Xe-138 5.80E:01 2.15E-03 Kr-85. 7.60E+00 2 .282E-02 Kr-87 9.80E-01 3.63E-03 Kr-89 8.40E-02 3.12E-04* The dose conversion factors taken from EPA 400R92001 (Reference 5.2) are listed in Tables 4.4 and 4.5 below.(I Radiological Release Thresholds CALC. NO. STPNOC013-CALC-002 E N E R C O N for Emergency Action Levels REV.Ic& Eeydqy Attachment I PAGE NO. 15 of 49 Table 4.4 TEDE Dose Conversion Factors Dose Conversion Factor Nuilide (rem per uCi*hr/cc)

Dose Conversion Factor (rem per uCi*hr/cc)

Nuclid6 t TOMWE"a~gum rffaffU¶4ýivivI.h.

VWý?GM 1-132 I 4.90E+03 I Xe-137 1 1OE+02 1-134 i 310OE+03 i Cs-134 6.30E+04 Kr-83m Te132 1.20E+04 Kr-85 1,30E+00 Rul03 r 1.30E+04 Kr30E+03 Zr95 320E+04 Xe-131m-Ce144 -- -. -4.50E+05

...Xe-133 i 2.OOE+01 Sr89 I 5.OOE+04~ ~ ~ ~ ~ ~ w .4 ..-A~aiI ___ ____62 Table 4.5 Thyroid CDE Dose Conversion Factors Thyroid CDE DCF Nuld (e e uCi*hr/cct 1-132 7.70E+03 1-134 1.30E+03 The unit vent noble gas monitor energy efficiency by nuclide is taken from Offsite Dose Calculation Manual (Reference 5.3). The values are relative to Xe-133 efficiency since the monitor is calibrated to Xe-133. Table 4.6 displays the energy efficiency by nuclide relative to Xe-133.

Radiological Release Thresholds CALC. NO. STPNOCOI3-CALC-002 E N E R C O N for Emergency Action Levels " Vl 1REV.f y .tlylorokCz5VfY Attachment 1 PAGE NO. 16 of 49 C Table 4.6 Energy EfficiencyRelative.

to Xe-133 Efficiency Relative to Xe-133 Nuclide Kj 89:. .,"*Thre s ip eltiv eficeny ailable for lKr-3m 'Assumtd .f~itther justifies the om isio:.It Table 4.7,Nuclide Half Lives Nuclide ..Half fe Nucide Half Life d...(ir?..

1-131 28 .0 23l eX137 .6.38E-02 1-134 8.077E3-01i::

Cs-134 1.8013+0.4

ý2 ý- I ýýAWnl 0TI119-Imm Kr-83m i-1 1.i831+00 ITe132."7.79E÷+01 Kr-8.3i kr-8'5 .:- 04Y IRUl3 .9.44E+.02.

Kr-88 2.84...00 Zr95 1,55E+03 Xe-131m " 2183E+02 144 6.82.E+03..... ...Xe-I337 1.27E+02 Sr89 1.21E+03 MI IVI WISE M5~T~h ~ 43* The half-lives are taken from Reference 5.15 which lists the input data used by STAMPEDE.

Radiological Release Thresholds CALC. NO. STPNOC013-CALC-002 C E N E R C O N for Emergency Action Levels REV. I Attachment 1 PAGE NO. 17 of 49

5.0 REFERENCES

5.1 Regulatory

Guide 1.145, Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants, Revision 1, November 1982.5.2 EPA 400R92001, Manual of Protective Action Guides and Protective actions for Nuclear Incidents, Revision 1, May 1992.5.3 Offsite Dose Calculation Manual, Revision 17, March 2011.5.4 TGX/THX 3-1, Revision 5, Westinghouse Radiation Analysis Manual.5.5 MC05 591,. Main Steam PORV Capacity Verification, Revision 1.5.6 NIST Steam Tables, 2011.5.7 OERPO1-ZV-INO1, Emergency Classification, Revision 10.5.8 OERPOI-ZV-TPO1, Offsite Dose Calculations, Revision 21.5.9 STP Calculation NC-9012, Process and Effluent Radiation Monitor Set Points, Revision 7 5.10 STP Calculation NC-90 11, CRMS Rad Monitor Setpoints, Revision 2.5.11 STAMPEDE Computer Program, Revision 7.0.3.3.5.12 STAMPEDE User's Manual 5.13 STP Drawing 6C189N5007, General Arrangement Reactor Containment Building, Revision 6 5.14 NEI 99-01, Revision 6, Development of Emergency Action Levels for Non-Passive Reactors--- -5.15- .:.-ITWMS Call- Nunbjer.10.000-10.987-Desigri-Documeart,-R evisipo 0. .-..- -6.0 ASSUMPTIONS

6.1 Release

lasts for one hour Per NEI 99-01 (Reference

.5.14), IC AA1, AS 1, AGI developer notes, the release should be assumed to last one hour'.For this to be true for the main steam line, it is assumed that the PORV is open for one hour. To calculate the most limiting case, it is assumed that the maximum flow possible is being released from the PORV.6.2 Nuclide mix Per 0ERP01-ZV-TP0 1, Offsite Dose Calculations (Reference 5.8) any unit vent release with increased RCS activity and no core melt should be calculated using a gap inventory.

It is conservative to assume an increased RCS activity and not within the intended scope of the relevant initiating conditions to assume core melt. Therefore, a gap inventory is used for all unit vent releases.Per OERPO I-ZV-TPO1, Offsite Dose Calculations (Reference 5.8) for a main steam line release following a steam generator tube rupture it is appropriate to use an inventory of 100 percent noble gases plus 0.2 percent iodine. Since a steam generator tube rupture releasing through the PORVs is the only steam :generator tube rupture scenario which would create offsite doses large enough to meet or exceed the EALs, this assumption is made.

Radiological Release Thresholds CALC, NO. STPNOCOI3-CALC-002 E N E :R C 0 N for Emergency Action Levels F .Ev I Attachnment I PAGE NO. 18 of 49 (6.3 Atmospheric Dispersion Ntf 99-01.(Reference 5.14) developer notes for initiating conditions AA1, ASI and AG1 suggest using the ODCM .or the site's emnergency dose Per OERP01-ZV-TPO1, OffsiteDose Calculations (Refesrence 5.8),. when. actual fiieteo0ology is not available, the default STAMPEDE values should be used. The default STAMPEDE values assume stability class D for aimospheric dispersion and a windspeed f 13.2 mph.Theýsev .aluel:'i

'vre sed s inputs for the caltclafion:-!..:

It is clear that STAMPEDE uses the same method for calculating.

atmospheric dispersion factor (X/Q) itli:ned in: section 7.1.1 of.: t ttacmht. However;'STAMPEDE.

does not follow the same logic in selecting the appropriae result from the three cacuations.

The STA EDE value printed in the r&ilts f6iind in attachment 3 isý consistentwitf the l4;gest of the three hand calculated X/Q values. This sug:sgts that STAMPEDB simply selects the iaig.st of the three X/Q values resulting in a much onservative estimate.

will deviat.e from the recommendations of RegulatoiY'Quide 1.45 ýand conform to the meth odoog! ' STA EDE uses.The close- po ity -of all re' leqpoints Allows for a single atmospheric.

dispersion coefficient to be bused. This as'osmptioh is also iade6"bySTAMPEDE..

6.4 Exposure

Pathways The dose conversion factors used intable 4.4 and 4.5 represent a summation of dose conversion factors for external plume exposure, inlhalation from the plume, and external exposure from (deposition, Becatise the dose estimations are used for implementing early phas'e'protective actionos, cdonversion factors using limi ted patthwayase appropriate-.., The EPA does not provide a dose conversion factor for K'-83m. Because the PAGs are based on EPA dose calculations, it is appropriate to oily use the nuclides for:which dose conversion factors are pvided d. Additi6nally, repres ents only 1.33% of the nuclide in;entory activity and its exclusion wold not signiplca',l' affect the final dose.'6.5 The release initiates one hour after reactor shutdown.

While a release initiating at reactor shutidown is ikely, significant decay of shortliyved nuolides occurs during the migration time. A"~ ~~~1 ee y,: I-s anud e

t. m0it9 -o*ainfm at the r~eception site. it is m tant for the thteshokl to not be calculated at shutdown as this would create ajvery high hshold :which would rioh be appropriate for releases which occur shortly after shutdowi.

one hour afer reactor shutdown is s to decay short lived nuclides and create a conservative threshold.....

Decay is incorporated for.bne hour from reactor shutdown 'as well as migration time. Half-lives are taken from Reference 5.15. Migration time is assutmed to be the reciprocal of the wind speed." ' : : ." " :. ." " ."". " " ." i " ": :" " '( " :

Radiological Release Thresholds CALC. NO. STPNOC013-CALC-002 E N E R C O N for Emergency Action Levels REV. I Evec- r Attachment I PAGE NO. 19 of49 7.0 HAND CALCULATIONS 7.1 Unit Vent Monitor 7.1.1 X/Q The atmospheric dispersion factor, X/Q, determines the change in concentration between the unit vent discharge and the dose reception site. This value is based on meteorological conditions and will vary with wind speed and stability class. The ODCM uses the highest annual average X/Q value at the site boundary which is 5.3E-06 sec/m 3.However, for an accident related release STAMPEDE is used rather than the ODCM. STAMPEDE uses real time, user entered, or default meteorological conditions to calculate the X/Q for a specific accident.

Default values will be used as inputs into the Regulatory Guide 1.145 method for calculating X/Q as described below. Default values are identified in section 6.0, Atmospheric Dispersion.

For a neutral atmospheric stability class, which is the default in STAMPEDE, X/Q values can be determined through the following set of equations.

Equation 7.1.1.1 X 1 Q Uio(3.7Eayuz Equation 7.1.1.2 X 1 Q Ulolry-cri Equation 7.1.1.3 Where X/Q = relative concentration (sec/mA3)7t = 3.14159 U 1 0 = windspeed at 10 meters above plant grade (m/s)dy = laterdl plume spread (in), a function of atmosptherc stabiity and distance, determined from Regulatory Guide 1.145 Figure 1 Cz = vertical plume spread (m), a function of atmospheric stability and distance, determined from Regulatory Guide 1.145 Figure 2 Ey = (M 7 1)Uy800m + oy = lateral plume spread with meander and building wake effects (in), a function of atmospheric stability, windspeed U 1 0 , and distance; M is determined from Regulatoiy Guide 1.145 Figure 3 A = the smallest vertical-plane cross-sectional area of the reactor building (mA2), taken from Reference 5.13 and shown below Radiological Release Thresholds CALc. NO.sTPNOI3-cALC-002 E-N ER C 0: N for Emergency Action Levels Rv.hproJ1yd Attachment 1 PAGE NO. 20of49 C Figure 7.1.1.1: Reactor Building Dimensions EL 241*- 0 S 6DtOESS.-... ..- .* ,* .0*4 .0" FUEL XR.OLI- i RE SUILOIIX LE L .". ': 6" if L %ATE5 ... .j' : ' z' " up M C I S "- ' .1E U .L 31'-L EL WO-C) FE XR R .C .E. , U. *4WFL 0 RING DUG " .L C-5-* I ...,.cu .-11 7iE Mo :I IR all : :-o t 2 ,-ELLEY. %-2.-S A!. Il1t. EUR.SdI-. -12T'-0" E L C -) 2 '- 2'" ... .f ..' -..'*" 'y ""'Re,.: .i Assuming the reactor building cross section to be a perfect rectangle and half sphere, the variables are defined as follows;U 1 0=13.2 mph. 5.9 ms-zoo .: a, 4.2 m Fly =(M .)..8., + .ay; l .-a'y 1 2 0 rn... ...A = (13'

• 15') + (ry ; = 31128.37 The three equations become;X .110-6 5.9 ( 0 4.2. * "31128.37 C; Radiological Release Thresholds CALC. NO. STPNOC013-CALC-002 E N E R C O N for Emergency Action Levels REIV. I&Ce~kne,- PWIAttachment 1 PAGE NO. 21 of 49 X 1= 3.568*10-6 Q 5.9(3r

• 1200

• 4.2)X 1 x 1 =1.07* 10-S Q * ,. * [(1 -1)Cy80om + 1200]* 4.2 To select the appropriate X/Q value, the first two X/Q values should be compared and the higher value selected.

This value is then compared with the third X/Q value and the lower of those two is the appropriate X/Q value. The appropriate X/Q is 5.39E-06 sec/In 3 for default meteorological conditions by the methodology recommended in Regulatory Guide 1.145.This calculated value is very similar to the ODCM highest average value of 5.3E-06 sec/m 3 which was not selected for use. Additionally, the value shown in the STAMPEDE output file at one mile is 1.032E-05 sec/m 3.This suggests that STAMPEDE uses the same methodology and simply selects the largest atmospheric dispersion value to remain conservative.

This methodology will be replicated and 1.07E-05 will be used as the X/Q..--- .2Nfilideilentoiiy

--As previously stated, a gap inventory is appropriate for this problem. The gap inventory is taken from TGX/THX 3-1 (Reference 5.4) which is used as the source term for STAMPEDE inventories.

The concentrations were then normalized so they could be scaled to the varying emergency classifications.

The values for the normalized inventory can be found in Table 4.2.7.1.3 Dose Conversion Factors As stated in NE199-01 (Reference 5.14) developer notes, the purpose of dose projections is to check if the Environmental Protection Agencies Protective Action Guidelines (EPA PAGs) have been exceeded.

The dose conversion factors provided by the EPA in EPA 400R92001 are used. These dose conversion factors account for external plume exposure, inhalation -from the plume, and external exposure from deposition and are listed Tables 4.4 and 4:5, andtkn f r -5=--1 2- P-A-4-m rR92-1 ferren -"_.2), The EPA does not provide a dose conversion factor for Kr-83m. This nuclide contributes 1.33% of the inventory activity.

The lack of this nuclide's contribution to the final dose will not significantly affect the outcome.7.1.4 Decay Time One hour of decay is incorporated to the monitor response due to the release initiating one hour after reactor shutdown.

Decay is also incorporated for the duration of the migration time, The total decay time is one hour plus the reciprocal of wind speed, or 1.07575 hours.

CALC NONTNC13CL-0 Radiological Release Thresholds CL.N.SPO~3CL-0.ENE CON for Emergency Action Levels .0-0 -"-t ,,y d",Attachment 1 PAGE NO, 22 of 49 (7.1.5 Dose Calculations The dose rate at the site boundary is calculated using Equation 7.1.5.1., n. 1.07575 D =-F

• 0.5 T 1/2 i

• DCF, Equation 7.1.5.1 Where"/) =d ose rate per hour at the site boundary X = atmospheric dispersion coefficient as calculated in section 7.1.1 F = Aut vent flow rate : C*"= conceitration of nuclide tiat the time of shutdown 1.07575- =the total decay time from' section 7.1.4 T11= the half-life of nuclide i DCFt* the dose conversion factor for iulido listed in tables 4.4 and 4.5 ,. .The total concentratiofl of the is varied to:find the d6se rate of interest.Beginning with an arbitrary release concentratioln of i gpi/cc the dos9 rate is calculated.

Since the dose is lineaYyyorrelated to. ooriceitration, the ireleasi conentration may be scaled to find thedoserate of interest.The Alert EAL s 10 mririm TEDbE or, 50 CDEB. Using the above method to calculate TEDE with appropilate, conersion fectors, a limitingtrlease rate of 2.33E406 jxCilsecirom-the unit vent results in 5.7 mre Tt Ijpsing the calculated release rate to0fmd with the appropriate c:nveion fact0s, the same h release resulthin a 50 reenr Thyroid CDE at thite Si budary. Thus, 2.33E+06 ipCi/sec is the imiting reease rate based on the 50 mem .r iThy'roid CDE EAL initiating condition.

The limiting release rate threshold values for the Site Area Emergency and General Emergencies are multiples of 10 and 100 ofthle Alert re ease rate threshold value.These calculations can be found in Attachment 2.*7.1.6 Monitor Response.The unit vent noble'gas monitor is calibrated to Xe-133. Monitor efficiencies relative to Xe-133 by nuclide are listed in ODCM Table B3-2. To find the monitor reading associated with each limiting release, the noble gas concentrations must be multiplied by the monitor response and summed. Table 4.6 shows the indicated response of the unit Radiological Release Thresholds CALC. NO. STPNOC013-CALC-002 EN E R CO N for Emergency Action Levels EV.I-llw--EvoepecLt Evty, d Attachment 1 PAGE NO. 23 of 49 vent noble gas monitor by nuclide and Equation 7.1.5.1 shows how the monitor response was calculated.

n Monitor Response C 1

• Rel Equation 7.1.5.1 Where Ci = concentration of nuclide i ([LCi/cc)Rej = monitor response to nuclide i ([tCi/cc)xe1 3 3 equivalent In the case of an Alert, the 2.33E+06 ILCi/sec release rate will read as 1.57E+06 lLCi/sec on the monitor. Kr-83m does not have an indicated monitor response coefficient.

Because Kr-83m is only 1.34% of the noble gases and does not contribute to the dose calculation, its exclusion is acceptable.

This again is a linear correlation and the SAE and GE scale by factors of 10 and 100 respectively.

These calculations can be found in Attachment 2.7.2 Main Steam Line Monitors 7.2.1 X/Q Since the atmospheric dispersion is independent of nuclide inventory or release rate and the close proximity of the releases, the X/Q value will be the same for a main steam line release as it is for a unit vent release. This assumption is also taken by STAMPEDE and outline in A~ssumption 6.3.7.2.2 Nuclide Inventory Pe r-UK'U R-Z WTPO 1, it the release path is the main steam line with a steam generator tube rupture, the nuclide inventory should be 100% noble gas and 0.2% of the iodine from the reactor coolant.The secondary steam concentration for noble gases and iodine after a steam generator tube rupture are taken from TGX/THX 3-1 (Reference 5.4). Values for the reactor coolant inventory are listed in table 4.3. All of the noble gases are used and the iodine concentration from the coolant inventory is scaled to total 0.2% of iodine in the. total coolant inventory.

These inventories are then normalized to one. These values are listed in Table 4.3.7.2.3 Dose Conversion Factors The dose conversion factors used are found in Tables 4.4 and 4.5, taken from tables 5-1, 5-2 in EPA 400R92001.

Radiological Release Thresholds CALC. NO. STPNOC013-CALC-002 E N E R C O N for Emergency Action Levels REV. I Attachment 1 PAGE NO 4of 49 7.2.4 Decay Time.One hour of decay is incorporated to the monitor response due to the release initiating one hour after reactor shutdown.

Decay is also incorporated for the duration of the migration time. The total decay time is one hour plus the reciprocal of wind speed, or 1.07575 hours.7.2,5 Dose Calculations Eqtiatioh 7.1.5.1 applies to the release from the main steam lines. The main steam line*

is used instead of the unit vent flow rate .for the value F. The main steam line flow rate was calculated in Eq uation 7.1.2.2 of the STAMPEDE CALCULATIONS section of this document as 2.79E+06 cc/sec.The Alert EALthrieshold is 10 mrem TEDE or 50mrem Thoid CDE at the site boundary (Table 4.2): 1jsis'"g e iethiof in Equaition 7,1.5.1 to calculate TEDE with the appropTiate" a condeenatii" at ti.me.of shutidow of 4.10 ýtCi/cc would resu t in 6.89mTiEDE.at:.th6e iti 6 b. fke "m .PORV was open for an p hursing jp.-gamq~ea tn~ line concehtration to dalciiathe.Thyroid ODE results in 50 mare hri CDE at the t b6 'd'ry Mb- -Thrsteam liin. cnc~fitrationsat the, imei. of fo r the-"Site Area-Emergency and " General Emnergencies of 16 and 100 of Aieirt.Since the correlation between release. concentrationt ad dose is linear, vaijes for steam line concentration at time f shitdown Aid 410 and C SAE and.GE respectively.

Both are also limited by Thyroid CDE:..These calculations can be found inAttachfetet 2 7.2.6 Monitor Response " .; .,." : ' Because. the main steam .line monitor is adjacent to~ he line, significant

.shieldingtakes place ~etween thesoce anmonitor.

STP calcuiation NC-9011 Revision 2 calculates a conversion factor for the main steam lines foranbe gas inventory which is i ncorporated into0the monitor readout. No monitor response needs to be calculated.

The concentration of the main steam line one hour after shutdown given a concentration

-of 4.!0 the time.of shuttd.wn is 3 .0 ýtC1/cc. This calcu1t-.iof is also founid in attarfinent

2. Additioihally, the iionitor tradings o'i'"tliet SAE and GE one hour after shutdown are 39.0 and 390 pCi/cc respectively.

These values are the thresholds for the main steam line monitor.(I Radiological Release Thresholds CALC. NO. STPNOC013-CALC-002 ENER N for Emergency Action Leoels REV. 1 Awwww r

• T ,ftAttachment 2 PAGE NO. 25 of 49 Table A2- 1: Unusual Event Emergency Calculations II I9A413+07I 1.40E+05 I 2.79E+06 '! I 5.OOE-02 Table A2-2: Input Values for Calculations

',+06 ob.ýc.O'sN]Radiological Release Thresholds CALC. NO. STPNoC013-CALCO002 for Emergency Action Levels REV. 1 Attachment 2 PAGE NO. 26 of 49 I I Table A2-3: Ca culations for Boundary Concentrations and TEDE dose due to Unit Vent Release* 1-131 1-132 1-133 1-134 1-135 Kr-83m Kr-85m Kr-85 Kr-87 Kr-88 Kr-89 Xe-13 Im Xe-133m Xe-133 Xe-135m Xe-135 Xe-137 Xe-138 1.1OE+05 1.12E3-03 2f76E-05 1 .50E+015 2.20E+05: 2,40E+05: 2-OOE+05.1 .30E+06 2.90E+06 3 370E+05 5.50E+06.7.80E+06 9.50E+06 1.1OE+05 6.80E+05 2.20E+07 4.20E+06 5.50OE+06 1.90E+07 1.S0E-'07 1.53E-01 2,25E. 2.05E-03 1 33E-02 2.97E-02, 3.78E-03 5.62E-02 7.98&-02 9.72E-02 1.12E031-6.95E-O 3 2.25E-Oi1 4.30E-02 5.62E-,02-1.94E01-O 1.~84E-01':

77E-.'05-, 06B-05 28E,-04, 33E-04 33F,05 39E-03 97E-03 40E-703.76E-05 b 6Eý-O3 39E-03~79IEýO3 549-03ý:1.QI-&03

"-.79&-08:.

i.01ýE-03'-

3.SEg 1.OI1H-03.

6.1 I-08 1,O1E-03, 3) 74E-07'1.4.1E-01 9.42&0&1 1.01E- 03 1.40E-06 1.01E-03 1.99B~-06 1 .OE-03 2.42&0-O6-41'.O1E-03

!2.79&40 1.OIE-303.

L,173EýO7.

1.015-03 ,T-'1.O1E-03 1.07E-06 1.O61 9-03mOE tLO1E:-O3ý

'.8Y-O6 1.Olt-ý03.

4.$9E-06 1 .93E1-+02 278-'2.3$E+00 2.79E-08, 2.03E.+01 , S.E0M8'--8.77&0-1 " 2,.61E'.0 6. 61E40OO- :4302-09.1.83E+00.

2.21,E-07 4.48E{-OO 6.27E-07 9.46E+04 9.42E-08 1 .27E+00 7.79&-07 2,84E+00O 1 .53E-06.d.:8!E+OZ-2.78E408 5.42E-O 1- :L,7-_h-07 I1.Z7E+02

5.57E-06 2.60E~-01 6.0.9&08 9.08E+00-1 1-.29E-016.

6.'3.W-02, .4,06F-11 2.,36E-O1 1 ..95E-07T 4.90E+03 3.IOE+03 930E+01 1.30E+00 5.1OE+02 1.30E+03 1.20E+03 2.50E+02 1.40E+02ý1.10E+02 7.20E+02 5.30E+04 4.90E+03 I.50E+04 1 .4! I1.-U.J 1.37E-04 8.11E-04 8.o9E-o5-3.70E-04 O.OOE+'O0 5.83E-05 1.22E-07 3.97E-04 1.99E-03 1.30E-09 1.36E-07 2.90E-06 1.11 _E-04 1.52E-05 1.81E-04 4.47E:09 1.40E-04 Jj ~Radiological Release Threihold&

CALC.NO. STPNOCOI3-CALCA)02 0 ENE O for Emergency Action Levels REV. Ii N_ _ _ _ _Z... , *Ltve~yd.2 Attachment 2 PAGE NO. 27 of 49 Cs-134 Cs-137 Tel 32 Mo99 Ru103 Rul06 Zr95 La140 Ce144 Ce-141 Sr89 Sr90 3.70E+01 2.90E+01 4.80E+00 1.22E+01 8.80E-03 2.90E-03 1.1OE-02 1.90F-02 7.40E-03 1.00E-02 6.40E-02 3.20E-03 3.78E-07 2.97E-07 4.91E-08 1.25E-07 9.00E-11 2.97E- 11 1.1213-10 1.94E-10 7.57E,- 11 1.0213-10 6.55E-10 3.27E-11 9.3: 7.3: 1.2 3.0: 2.2, 7.3 : 2.7(4.7!1 .8~2.5i 1.6 8.01 E-09 E-09 E-09 E-09 E-12 E-13 E-12 E-12 E-12 E-12 E-I1I E-13 1.0113-03 1.01E-03 1.01E-03 1.01E-03 1.01E-03 1.0IE-03 1.01E-03 1.OIE-03 1.01E-03 I.01E-03 1.0IE-03 1.011E-03 9.42E-12 7.40E-12 1.22E-12 3.1 1E-12 2.24E-15 7;40E-16 2.79E-15 4.83E-15 1.89E-15 2.54E-15 1.63E-14 8.15E-16 41.80E+04 2.60E+05 7.79E+01ý.62E+01 9.44E+02 8.84E+03!1.55E+03 4.03E+01 ,.82E+03-7.777E+02

,1.21E+03 2.50E+05 9.42F-12 7.40E- 12 1.21E,-12 3.08E-12 2.24E-15 7.40E-16 2.79E-15 4.75E-15 1.89E-15 2.54E-15 1.63E-14 8.15E-16 b.i3Ub+-U4 4.10E+04 1.20E+04 5.20E+03 1.30E+04 5.70E+05 3.20E+04 1.10E+04 4.50E+05 1.101E+04 5.00E+04 1.60E+06 5.93E-07 3.03E-07 1.45E-08 1.60E-08 2.91E-11 4.22E- 10 8.93E-11 5.22E-1 I 8.49E- 10 2.79E-11 8.16E-10 1.30E-09 T~~T1~~Dosw~5.7EO I ;

Radiological Release Thresholds CALC.NO.S'1NOC013-CALC-02 0 for Emergency Action Levels RV- I Attachment 2 PAGE NO. 28 of 49 Table A2-4: Thyroid Dose Calculation for Unit Vent Release 1J-1. I 1-132 1-133 1-134 1-135 2.78E*-08.

2.7918-0&5.40E-08 2.6 1E-08: 4.56E-,08 1 .30P-406 7.70E-1,033 2.20E+I05 1 30E+03 3 .8OE-ýO4ý.3,61E-02 2. 15E-04 1. 19E-02 33.39F,05 A.1 71i A"ýTable A2-51 Unit Vent Monitor Response to Nuclide Inventory Kr-85m Kr-85 Kr-87 Kr-88 Kr-89 Xe-l31m Xe-133m Xe-133 Xe-135m Xe-135 Xe-137 Xe-138 7.3 )3E-04),33E-05[.9E- 03[-.97E-03 l.40E-03ý.76B-O5 (.71E-04.5.55E-03 1.06E-03.319E03.79E"03.54E-03 4.48E+00 9AOE+041 1 .27E400, 2.84E+610-5.1 OE-02 2.g3E+I02'-

5.42E0-I-Ol 2.60E-O1-9.OSE-4-OO 6.38E-02 2.36E-01 6.28E-04 ,1.9 9.33E-05 2.4 8:.03E&04 2.8 1.54E-03 2;3 3 OOE-09 2.8 2.76E-05 0.015 1,69E-04.

0.-14 7.38E-05 -O.O-42 1.28E-03 2.5 9.15E-08 2.8 2.41E-04 2.8 Monitor Reading: 0.00OE+00 1.19E-03.2.24E-04 2.25E-03 3.55E-03 8A40E-09 4.1.3E-07 3.21E-03 2.56E-07 6.74E-04 (uCi/ec) (uCi/sec)

Radiological Release Thresholds CALO. NO. STPNOCO13-CALC-002 W E N E R C O N for Emergency Action Levels REV. I-&C--EM ýPWOCE Attachment 2 PAGE NO. 29 of49 Table A2-6: Input for Main Steam Line Release Calculation 37Eb+Ub JjjS-t-79uI-U 4.U-'Tnhl,- A') zTzliici fhr Rmnincbinr rnni".P~ntmsfinwm nnth TP.TDFAQ.

fAn phi tn Mnin qti-m T.int- PAnz k1.-2 1 1-132 1-133 1-134 1-135 Xe-131m Xe-133 Xe-133m Xe-135 Xe-135m Xe- 137 Xe-138 Kr-83m Kr-85 Kr-85m Kr-87 Kr-88 Kr-89 0.1 U1-UV, 8.61E-02 1.OOE-01 1.86E-02 2.73E-0 1 2.80E+00 2.40E+02 4.20E+00 7.60E+00 4.OOE-01 1.60E-01 5.80E-0 1 3.70E--01 7.60E+00 1.50E+00 9.80E-01 2.80E+00 8.40E-02 2.26E-04 3.19E-04 3.72E-04 6.92E-05 1.01E-03 1.04E-02 8.90E-01 1.56E-02 2.82E-02 1.48E-03 5.93E-04 2.15E-03 1.337E-03 2.82E-02 5.56E-03 3.63E-03 1.04E-02 3.12E-04 9.27E-04 L31E-03 1.53E-03 2.84E-04 4.14E-03 4.26E-02 3.65E+00 6.40E-02 1.16E-01 6.07E-03 2.43E-03 8.82E-03 5.62E-03 1.16E-01 2.28E-02 1.49E-02 4.26E-02 1.28E-03 2.9853E-05 2.9853E-05 2.9853E-05 2.985'E-05 2.9853E-05 2.9853E-05 2.9853E-05 2.9853EH-05 2.9853E-05 2.9853E-05 2.9853E-05 2,9853E,-05 2.9853E-05 2.9853E-05 2.9853E-05 2.9853E-05 2.9853E-05 2.9853E-05 2.77E-'08 3.90E-'08 4.55E 2 08 1.24E-,7 1.27E106 1.09FA04 1.91E-06 3.45E406 1.81E-07 7.26E-08 2.63E-07 1.68E;07 3 .45E 4 06 6.81E-07 4.44EH07 1.27E406 3.82E-08 1.93E+02 2.38E+00 2.03E+01 8.77E-01 6.61E+00 2683E+02 5.42E+01 1.27E+02 2.60E-01 9.08E+00 6.38E-02 2.3611-01 1.83E+00 4.48E+00 9.40E+04 1.27E+00 2.84E+00 5.10E-02 2.76E-08 2.85E-08 4.39E-08 3.62E-09 1.1OE-07 1.27E-06 1.07E-04 1.90E-06 1.96E-07 1.67E-07 6.10E-13 1.12E-08 1 12E-07 2.92E-06 6.81E-07 2.47E-07 9.79E-07 1.71E-14 5.30E+04 4.90E+03 1.50E+04 3. 10E+03 8.10E+03 4.90E+00 2.OOE+01 1.70E+01 1.40E+02 2.50E+02 IAOE+02 7.20E+02 1.46E-03 1.40E-04 6.58E-04 1.12E-05 8.95E-04 6.22E-06 2.1 5E-03 3.23E-05 2.75E-05 4.17E-05 8.53E- 11 8.04E-06 O.OOE+00 3.80E-06 6.33E-05 1.26E-04 1.27E-03 2OSE- 1 1.30E+00 9.30OE+01 5.101E+02 1.30E+03 1.20E+03 IToalose 038E-~*Release Constant = X/Q

• duration

• release ra MNO. $TPNOGO13-CALC-002

...N.,CO ,N for EmergencyAction Levels -'RV.-_.hm e i Attachment 2 PAGE NO. 30 of 49+ablA2-8:

Line ReleaseThyrid.Dose,.Ca!culation 1-131 1-132 1-133 J-134 I-:13 5 2.76E-O8 2.95E-08.'A;39&-08 3.02F-O9 1.1OE-07 7.70B+91, 2.20E05ý2.20E-04, 9.66E,-03 4.7 iE-6 4.26M-3-I for ERadiological Release Thresholds CALC. NO. STPNOC013-CALC-002 r N for Emergency Action Levels REV. 1 0 -vwp~1-d~

Attachment 2 PAGE NO. 31 of 49 Table A2-9: Main Steam Line Reading at Release 1-131 1-132 1-133 1-134 1-135 Xe-131m Xe-133 Xe-133m Xe-135 Xe-135m Xe-137 Xe-138 Kr-83m Kr-85 Kr-85m Kr-87 Kr-88 V.4 fr1-03q 1.31EB-03 1.53E-03 2.84E-04 4.14B-03 4.26E-02 3.65E+00 6.40E-02 1.16E-01 6.07E-03 2.43E-03 8.82E-03 5. 6.2E-03 1.16E-01 2.28E-02 1.49E-02 4.26E-02 I I ý Y.5 ,-t-JL 2.38E+00 2.03E+01 8.77E-01 6.61E+00 2.83E+02 5.42E+01 1.27E+02 2.60E-01 9.08E+00 6.38E-02 2.36E-01 1 .83E+OO 4.48E+00 9.40E+04 1.27E+00 2.84E+00 1C Y.Z.J-U4 9.77E-04 1.47E-03 1.29E-04 3.73E-03 4.25E-02 3.60E+00 6.36E-02 8.04E-03 5.62E-03 4.65E-08 4,67E-04 3.85E-03 9.90E-02 2,28E-02 8,62E-03 3.34E-02 I 'CAn An CALC. NO. STPNOC013-CALC-002 Radiological Release Thresholds 01 ENER C-O N for Eiriergency Actiol Levels "REV. I" Attachment 3 PAGE NOý 32 of 49 DRILL.' TA"EEImrh pbe Ifo C~ouU96 Wdear .ibLa12&e V-Sa kk, am 3#L~U~e~s~m*Ere~tuo14d~6ucIs 1ztcrmiu~~i1ef 09W I215-U Ikt~e5eI~'m l Ilu DPr~ I10Gu~iE&C C Xa .14O5 4.111& Y7E+4XB 2 9r4 5 200~1S2~.'*!-133: ~~XJ~B -S-1S~: SI2Be003 43*-134ý ,~D IA-14k:i' 53ME45 TE-23Z IME-00 fr~: 313W-dG X~~-24& 2~~4O3'1 1247)W13321t40 PM C CALC. NO. STPNOCOI3-CALC-002 Radiological Release Thresholds REV. I C EN ERCO N for Emergency Action Levels r.Szceeýn-Enerypm~t Eviyday Attachment 3 PAGE NO. 33 of 49 DRILL DRLSTL e, Imor ion DuWlh&127/21M315.24 l.k~r RiM: UnhVent Alrt ID.75 1020 Diflhmu tamilor)05 Iff 2.0 S2 75 (miles)05 1.0 5D 7.5 2w0 Plaea T~sn tM=e 023 V34 0:45 1:31 CELQ Vdies LFA0325 L0326MO 323S007 ISIIE-OO?2A3OZB 73I5E-MD 2.441E4Xf IKIM= w, anm tLOoW am~0-nix..M..... 'u CDR 0216 aUm am0 am'D 0.000 Dine am0 0-n-7 0124 zon Immcrdowfetb Hod7 rnAla paS Liia (rnaI amc 0.000 0Mc (zs3 (rMA MA0 0M5 Ono.8am 12AT72f23fl231#J CALC. NO. STPNOCOI3-CALC-002 Radiological Release Thresholds N.E. R CO for Emergency Action Levels rEV txim-&e q.gdet IVey dqý. Attaclhment 3 PAGE NO. 34 of 49 C DRIL ~LtZt ILL" RESLTS"l6 2d'T ow

....-Wl kiy 13aihr Re "'a1te, 1.I9 uCUE STAED ThOR 0- 0 Q1O '.002 01101 OLM0 A "Geeral Eme6re ncy R :e.qiiires aPirotective Aif 0"." Recommendation EVACUATE ZONE4) I SITJER fiN PLACE ZONE<, .. 2 .AFFCTED DOWNWINDSEUOB

&I R, A, B Al .".o-"". Zones .Inor A.-".ndl Moitr.A :.di .S '"tion Based amikDos-e Rate Projtictio"ýf>

3 mreinibfr (im-rnesion Whole. Rody KNobe Gas Gamma) aith Site Bonndaq Mie -r 135 niinilesor longeri~ thMe Trgency Ckd~iflcation IntaigConditionRAT (A .~)bhas been Met.C PhYEM tAJh BY: REufl'E By, Rad Ihina~ert/ffiologkrm Dirnetr.12)1/2013 324:M PM Datdfliine 12J1712013 3:24:23PM CALC. NO. STPNOCOI3-CALCOO02 Radiological Release Thresholds REV. I E- N E R C O N for Emergency Action Levels .I veyqay Attachment 3 PAGE NO.35 of 49 DRI,.LL. T EDY User Supped. normation

_DILL DhIbW:m M2fV2D3 0754.Comanft UrwDinr Stemn~~jtnriAlrt Ma. ."l u hfnr b --1Itnrcbo&A D~sh Ing4k: Qunimliddidoty:132 mtuh£~mkpdudkw1 5frnu 155 gra U"kwI2seheStshlIyas S&fiiiythss:

Ded!I&tdued fC.tGuSkxmrn kRkns SAdfffy &Itiir0u sftsnllowhf tDSvfnbr... ..rShu T. ...Rd~e~mSbat Diaflanam:

12 030:54 Niidit bkfrrc XAlieAW+I'anMM biM a ~ IONS u D2%A C~k~tedN~flZ~~rtas ztc3.15*1*?Ciflecr NODLECAS wacdi& RUtleC ZOtDlE Nudi& u¢i/sar* PATWUIMATS 7fwta oBisec*431k Xr-fl: XZa-13m Mt-33-1t Xp-338-1.1t-0400 3,43E+WS 25-%5M4 429MCG 121*icO I; nfw&Iwo-h IMR&W IM+CWB 12M-G Jr-fl I-rn I-rn: 1-131: 322E+=N 422E+003 123E*OM Ce-141: La-hO: Tra-99a O02m+=0 00E4030 0.OflE*WD.Oapnjom ODE3ac Aos-Cm 12fl5120 75519 AM r ,,. ,. ~~T.%flflt~

fliT CTAA4 I LALA~. iNU. 0 ir1NU~Ui.,-ui.-U~Jh

0. E NE RCO N I CALC. NO. 3 1 PINUCU L3-LtUA,-uvz Radiological Release Thresholds for Emergency Action Levels Attachimtent 3 REV. I PAGE NO. 36 of 49 C in~Phm duý 0l074 Bnir .h: mtS~l~(uziI.~~0.5 LO 5.0 1.5 10.0 2fl0 DI~i~ce cL5 1.0 141 5.0 75 10.0 20~msIii~z (u~Uh~)05 1.0 10 5.0 7.5 1410 2410 034 axmo.bi .&00~w~o UU~gMu -.2 ....31Sl437 (LOW:-00-0'.7..0.0m. 609 1*&6*e CBS~.*3*i2ift='UPS tow0 023 Ci 12ITUM137'%~4A2AM C

CALC. NO. STPNOC0I3-CALC-002 Radiological Release Thresholds

() EN ERCO N for Emergency Action Levels REV...JelcdPm-,p)ec-t.E-y day Attachment 3 PAGE NO. 37 of 49 DRILL sTAMR-1"D DRILL-C~ic~Il~ii~nICmlehd RESULTS WindVelofti 13.2 ifi~fr D-Inan~ireim ISO 3.L-died uf PrQa'icftai STAADUM I mie I iles 5 Mi" l1miles 71MD 007O 0.003 Mol 0J CDE D&053 0.017 0.004 -1 R aJ2dC&Mf~ioi~

reh3%- LkG~A General Emergency Requpires a Protective Ac~tion Recommenda~tifin EVACUALTE ZOM~(S)i I SHELTER~ IN PLACE ZONFA5) 2 AFFEC-TfDnOGWN-W1NDSEC-TOPS- -R, A. B ------ .-AE Ikamaimiqg Zones Go Inbors And Mcnitor EAB R~adio Statim Based ona Dose Rate Projection of> 3 mrem/br (Immerion Whole Body Noble Gas Gaunna) at &be Site Boundary (I Mile) for 15 miimttesor longer Ilie 2mergency Cbsaification Initiating Com][itionRAI (ALERT) has been met.PMTOMRMD BY.I'IEWEED BY:~12(19013:

755:14 AM DatiTim*I --1W11013 7:34-43AM CALC. NO. STPNOCO 1 3-CALC-002 Radiological Release Thresholds CALC. NO.ITPNC013CAL-00

ýE N RCO N for Emergency Action Levels: Attachment 3 PAGE NO. 38 of 49 D IL SAPDU.Sup fmai :DRILL C~ma*xd1-13hodt 332uithM 0uufkii~ke 130 &guM.e-dcrl unity .I'm ShttwytIMAM X&bzlhrttbl -etldn 2A003H7 trltsa N3ursr-bIatfLMkUrl~llil-L PARWITXIEM Nidi it/Sec.Ceflia-Ut-23.COO Xs-1t1. 2.3O3: re-W.2 3135M340 ,.J 1ll7AM201 3 3 315: WM..(

CALC. NO. STPNOC013-CALC-002 Radiological Release Thresholds EN ERC0 N for Emergency Action Levels REV. I ieeria.-F.ory plecI Ewyday, Attachment 3 PAGE NO. 39 of 49 DRILL STAtDRAuII -oMnI I Thwflmc. =amlt0 1525 Te N=Ws: U1&IW m"l AMa Cmaskt..... ~Iflnelfur am .... .....IJUOaisL (uilbs)0.5 LO 5.0 7-5 1&00 Ph.,twval 7ima@kcounimsuo).

I.23 OAS 131 CHMtQdah Lmo-)00 I.00IE406 L541K-W VHUQ Drm 143dR-M 3.,g1514.00 Z4413-00 MU00-M0 035 L.0 2.0 5.0 7.5 100 I mesuMpva Dos 7S IompxmaroWkeo Body can tor 023 0.0012 410M 8M1 3ME Iwdiz CUR cml2 M.176 OAS OL041 tiem (miles)03 1.0 510 7_5 100 aim U.00 OMI Om1 OM00 Q.002 am00 1394 03510 O.176 0.541 0014 am11 U4'17,2013 3-.,221 PM* I* I CALC. NO. STPNOC013-CALC-002 Radiological Release Thresholds ENE CO N for Emergency Action Levels g fr..d*y.Attachment 3 PAGE NO. 40 of 49 DRILL :~cninDRILL S__.:- ','M ""T, ' S.'"., ' -.'.R!( ;: S *.. .i:'? :i :....: :!~OfbklhInjedou (i wiles EM! ftOO 00.5 mf~n 3.OL041 ReleMe Rate: Ii 9 B4O0?uCi&~e E-7 I DAN2*D 0.5t0 A General Eoergency l 4 equfres a Potectivwe Action RecommendLatio..

EVACUATE ZONE(S)..

1: SHMtTERIN~

PLACE ZON(S 2 AFFECTED DOWNWIND .SETObRS:.

Rb ARB All Remann Zwies Go kidbmi And McititarEAS Rdio Station: Batd oa iteBewday ( Mi~Dose Prajacdian>

0.1 remý IBDE and("r 05 rman Thyroid CUEtu EmretCasbnli atin .ditionlt~l

'(SITE AREA EMERGENKCY) bat beeninet E'JEIED BY.Radnoaelitllo~a]Dinetor DatdTime 121017201 3 3I22I PM ICALC. NO. STPNOCO013-CALC-002 Radiological Release Thresholds ENERC0 N for Emergency Action Levels REV. I vhrydoy, Attachment 3 PAGE NO. 41 of 49 DARMDE -User Supped lufermationILL DPJLmL nVL*7..3 /ul D Dutaltfmr 172J212c3 23 Uek~wfae StsnbaUStAks CMMabma S etlnldhnnta QaMM~idkluaduhitr U2msthr tker-SaectdM~luilht1ss SbUrdyhwa "D-Nehat SfrnnkflfrM 454511 uCcwh StmnOflatez:

115 iffhuur I , ý3mmvqýRadez Start 1&tiikt 3*lure:s Na~e Gaf +Iet..... -... .... .......

U:2&13Wf213 15:21 LOD kamn Ifr eaapeet+ l *e (iNl .uCAS Ki-taM&C-35t X-ab: KT-lfl-t ss-na Trz-na XZ-l23M 1.149+Mt--

314M+W16 S+/-31z+4Th 255+1M+0 127E+406 121U+400 134E+004ncssec 3-MIl 3&01+.-13"Uk 323B ,-13-m: 4104 1-M34 4.22E+=3 1-135: .145*00 PARFUUATI Macy& n/SEZ Ia-14: Ea-b-99: Fal-IftS SrrIy-g fr-SI: 20310ER 0.031440 01U24400 0401+100 0OCCE+10W 0.DIEW40 0.00+000 0011+000 0.0as400 0.02+400 2021*0CC 0025+"2-I 112212033 32:2T03%i CALC. NO. STPNOCO13-CALC-002 Radiological Release Thresholds E0 .NER C N for Ef rgency Action Levels __REV. I.xCe. ..a-E.i:J1-16,6 Attachment 3 PAGE NO. 42 of 49 D STAMLvhm7DB3 A2s~~k~aii DRILL 1UMim7.r)J IMflMOU 11of 2D.75.20f 00-:002 034 05 1.A 2.0 7.5.05 75 20D CuLOO fizz1 0.001 01CM (175.025 (1006 OMI CMD ~ 0-M0 C 1UM17A033 32RM53-14 C CALC. NO. STPNOCOI 3-CALC-002 Radiological Release Thresholds E N E R C 0 N for Emergency Action Levels RV. I£xaone.r-EVeqypra~.

pf#vday Attachment 3 PAGE NO. 43 of 49 DIRIL RzSt=7 1 t 3 Page2 DRILL..M.. I cuhiacus Cxmsete ESMULTS WindVela-tys S2lminii"'WidDiretdiam 190...... .......STAMPEDE Reineae Rant 1.20E+/-0$nUCse Ofhfite Dose Projeetion (mm) i I =11; 2in Smiles ks I a ite""IEDE 0p 0:025 U06I 0.002 CE D.506 0175 0.01 0.013 Projecte~dd Afinamofrelease:

1.0 hour0 days <br />0 hours <br />0 weeks <br />0 months <br />st

A General Emergency Requires a Protectiw Action Recoininendation EVACUATE ZONE(S): I SHELTER IN PLACE ZONE(S): 2 AFFECTED DOWNWIND -SECTORS: .R A¢ B- -All Remrning Zones Go Indoorr And MonitorEAS Radio Statioan Based on a Site flomndTy (I Mile) Dose Projection m 01 rem TEDE aad/or D5 rem Thyroid CIME the Emergenzy RSI (SITE AREA EMERGENCY) hia been met p , I PEFORME&D By3r MIMED~ BY-12#IRO13 3:25:02 PM DWlThUZ a.12titlel 3l:203 PM CALC. NO. STPNOCO I 3-CALC-002 Radiological Release Thresholds E N E R C 0.N for Emergency Action Levels FBV1y r~da Attachment 3 PAGE NO. 44 of 49 m .flzfrJflSK 1211Y12013 1525 ThrNn~Uch1~uttkD CuiilsCdtI D......b-, .. :. :..

.1 mit.hr ..* u~a~t ~ '&...e'Thrs a :kct -':h': .: : "lt 5j ;..."

• d.i" 2 ' a"" " ""e" ' a " .' .Th t a~tut ..e~t sk a ......E4O .I -. .ti .. .Enarlmv m at/T I2C I.2Infl:..nI.S..&2....lt 12/2ZfI 5 :2"1 .EklkasffiAAfrU fbf I.- banns = ....,hh...C JJ~2cnjdedKOEhECASU1tZhratO I19B-OO3 nOikec riu ... *: ..IMM.. Is 1 ., M.i..- ..: E j... fi* ..: .l ThI .t 04001 x,-nt ' a-ol 1-132: 1-133: uGi~s 312fl15 5.125*005* m" "'" ...". C" 1 ".0 Cdfr-1l4t.

2 .*2.*Ce-141 S4tO Srl-1It- 13210,01.-3541 TZr-lit zn..oai 12Ul2,20133."7PAM

.

CALC. NO. STPNOCO I 3-CALC-002 Radiological ReleaseThresholds NENERCO N for Emergency Action Levels REV.1£Ecefra-Ev&erDid

ty,'da Attachment 3 PAGE NO. 45 of 49 DR]L STAMPEDE .Resilts Whormatui D IL z~hm RoaumMJ1 9MM1 buN.e: U1~dmL ciiMM111i 0.5 1.0 2.0 S.C.75 20.0 Dis1tiim 0.5 1.0 IC 5.0 7.5 10b0 75 20.D L0~2 005 0~3 0.4 131 L03MK-ow L5412.O07 7.373E-00 3144E.00 2.410M-00 Iza'Sozmg Mku Body 033222 0160 00103 Iauxm~iaa WbBoe Bary nobbj go am r (rw 0"33 0.016 0.610 0210 1.762 OW?7 0214 416170.135 0115 0.411 ow.2 0.U14 0.017 OL135 0.036 am005 12117.=13MZMA50.

CALC. NO. STPNOC013-CALC-002 N Raiological Release Traesholds

______________

E N E RC O N for Emergency Action Levels RFI. I.sy Attachment 3 PAGE NO. 46 of 49 WiauZinsmt.

Releae POe: 1.I9E+VOuCtaee TlOE 0.631 , (21 "5051 e.017 AGeneiral Enaae-rgernyl1iquires a Protective ActinRcm edto EVkCUATE ZONE(S) 1.,1 SHELThR IN PLACEZN(M 6 11 AiFFETED DOWNW IVM SECTQE& R, A. B All Rta4itg Zoesn Go Indoori And MonftoiEAS RAdi Sialtina-Based on a Site Boimduxy l l i .le) Dose Proeetioun 1 matfTlflE andlor 5mniranyrid CDE te EbrSrgey ClIAssfationlnit&h incnita l (GjENERAL EMERGENCY)hRbaa ee 'met (C .!rmrnxiub E-1212013 3:2(13 MI Datalfma Nnww1az lIwl cirecrr uaWIUWna 12JI1 .0 '33:26-25PM (I CALO. NO. STPNOC013-CALC-002 Radiological Release Thresholds RAEV. NOI TNC13CL-0 10 E E RCO N for Emergency Action Levels REV R

.Every day Attachment 3 PAGE NO. 47 of 49 DR]iLL STAWED srtupid InformationDIL D 1drmp- 12013 15-0 11w Namn:- Stumnmi QmiAa Cnmmaut: *

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MENER CON Radiological Release Thresholds for Emergency Action Levels.Attachment 3 CALC. NO. STPNOC013-CALC-002 REV. I PAGE NO. 48 of 49 DRILL~~~ 3 !RdEXsjtst IIIrn DRL 2.0 05 20.0 75 100 2.0..05 10.0 20.0 3.5EME40; 3Z15,0ýVM2 1664l100 737S3k67 0.45 335DWIE. 14MIE-M0 CTMWz .j A .l r)0.1530241-747 0.04am 0.5347 0.022 0.34bilk1 0.040 A07 1022 003402 0.013 ~~0W2 .3 (o 1171/fr133.0391&.{

C CALC. NO. STPNOCO013-CALO-002 Radiological Release Thresholds E E C O N for Emergency Action Levels REV. I

.£.veyi~do Attachment 3 PAGE NO. 49 of 49 DRIL STMEERslsInform afionDRIL Cak-ubdi=u CoQmj~he RESULTS iindDiretdcmw 30 off~ie Moae Prefrtio (rom): I smile CMe Ptojded&n~uofeiae:

D hou 101afes 0.L22 0.134 miraes 0.254 11747 SLmik 0.063 DM.47 A General Emergency Requires a Protective Action Recommenrdation EVACUATE ZONE(S)- 1,2.SHELTEP, IN PLACE ZONE(S): C 12 AFFECTED -DOWNWIND SECTORS: -R, A,-B -Al Remairing Zon Go Indoos And Monitor EAS Radio Station B aed on a Site Bowndary (I Mile) Dose Projection

-I rern TEDE =dor S rem Thyroid CDE the Emergey ChssiEfation Initmin Condition RA (GEINERAL EMERGENCY) has been met PXErORMED BY:

&"30:48 P.DWWTfnk~n RLMVMw- BY&legealDkiem nag 91-ft; I 12/17.f2015 3:-3OSDFM STPEGS UFSAR The partioulate channel is used as part of the Reactor Coolant Pressure Boundary (RCPB) leakage'detection system. The sensitivity and response time of this part of the leakage detection system, Which is used for monitoring unidentified leakage to the Containment, are sufficient to detect an increase In Jeakage rate of the equivalent of one .gal/niin within one hour. Elements of this monitor, including thb Indicator mounted in the RMS CR cabinet, are designed and qualified to remain fRintoional following.

a Safe Shut-down Earthquake (SSE),, in complianoe with RG 1.45. Further Information on the RCPB leakage detection system is presented in Section 5.2.5.11.5.2.3.3 Unit Vent Monitor: The unit vent monitor samples the plant vent stack prior {o discharge to ffie environmnnt and monitor for particulates, iodine, and noble gases.The unit vent particulate and iodine monitor draws representative air samples from the plant vent stack. via isoldnetie nozzles in the stack, and .directs them through a moving filter paper monitored by a shielded beta-sensitive schaillation detector.

The sample stream then passes throuagh a charcoal collector,.

vhere collected Iodine is monitored by a shielded gamma-sensitive scintiilation detector, The sample is then returned to the vent sltack.A separate wide-range gas monitor is provided for the unit vent. The monitor. has two isoldnetio nozzles for sampling.during both normal and accident conditions.

The stack samnples pass first through a sample conditioning unit which filters particultes and iodine and may' be used to take grab samples, The samples then pass through the shielded detector assembly, which uses thre dotectors to cover tDe complete range required.

The low range detector uses a.beta-sensitiveplastic sointillator-photomultiplier (PM) tube. The mid-range and high-range detectors uwe oadmium... telluride (CdTe), chlorine-doped, solid-state sensors. This wide-range gas monitor satisfies-the requirements of NUREG-0737, Item UY,1 for provisions for sampling plant effluents-for iodnes and particulates and for noble gas effluents from the plant vent.11.5.2.3.4 Control Room Electrical Auxilia, Building Vntlion. Monito The CR/EAB ventilation monitors are Class lB monitors which continuously assess the intake air to the CR for indication of abnormal airborne radioactivity concentration.

Each monitor assembly is powered from a sepgate. electrical power source., In the event of high radiation CR emergency ventilation operation is. initiated (Section 7,3.2). Failure of a monitor is alarmed in the CR.-Each monitor assembly Is compriset of arecirculation pump ,d4ensihr, .biiiiIl~fi6n d-bedr.four-pi lead shielding, check source, stainless steel sample gas receivtig chamber, and associated electronics, 11..2.3.5 Con,.denser Vacuum Pum= Mgnitor: Gaseous samples are drawn through an off.-ltne-system-by-a-puiip- -fTom-the-disehargo-of-the-wigt-ur-n-pun-p-exhaust-header-of-the-conde'nser.

This channel monitors the gaseous sample for radioactivity which would be indicative of an SO tube leak, allowing reactor' coolant to enter the seoonxdary side fluid; this monitor complements the SGBD mtonitors hi indication of a SO tube leak.' The gasebos radioactivity levels are monitored-by a single detector in a manner similar to the unit vent wide range gas monitor, 11.5.2.3.6 Spent Fuel Pool Exhaust Monitosm .The SFPB monitors are Class 1B and are.identical to the CRIBAB 'ventilation monitors described in Section 1.1.5.23A4 except that they sample the exhaust from the FHB. In the eyvent of high radiation the monitors initiate emergency operation 1.1.5-1.1 Revision 1,4 STPEGS UFSAR]1..5.2.5.1 Gaseous Waste Processing System Inlt Monitor: The GWPS inlet monitor employs a ganmma (Nal crystal) soiftillator/photomultiplier tube conibination to itleasure the radioactivity level of the waste gasos entering the. GWPS. The monitor is used In conjunction yith the GWPS discharge monitor to measure overall effectiveness of the GWPS, 11,5,2.5.2 GWMS Disoharge Monitor: This monitor is similar to the GWPS ihlet monitor and Is installed upstream of the QWPS discharge valve. Upon detection of high radioactivity or monitor failure, the GWPS discharge valve, FV-4671, is automatically closed, 11.5.2.5.3 Main Steam. Line Monitors:

Each MS line is monitored by an ATL monitor consisting of.a Geiger Mueller (GM) tube detector and an ion chamber detector with overlapping ranges. The detectors -are shielded by 3 in, of lead.The monitors are desigtied to monitor gross gamma activity in the steam line and provide a basis for detertmining possible atmospheric roleases from the MS power-operated relief valve (PORV), SG safety valves, and/6r auxiliary feedwater pump turbine,.The monitors provide a dose rate range'equivalpnt to 10.1 to 103 ýLci/om 3 xenon-133, BEased upon core invent6ry, the ratio. ofxenon-133 to other nuclides in the fuel can be detemined:

In order to obtain the above concenlrations of xenont 133 in the main steam line, a large primary-to-secondary leak must be present coinoident with a large amount of fuel failure. The presence of xenon-I33 indicates other radioactive isotopeý are present.Using the relitive ratios ofisotopes present in the MS line, a computer model for determinati on of dose rates from these isotopes, detector response curves, the thickness of the MS line, and thde geometry of the M9 line relative to the detector, the dose rate line concentration is-obtained.

The quantity of rddioact-fve effluents released is obtained by mintiplying the ienon-13.3 equivalentfMS line concentrations, by the isotope ratio titles the. steam release rate, These detectors:

are safety-related Class 1B and meet the requirements of RO 1.97 and NURBG-0737, Th5 S e~atih 0 Gbl6 Ml t --v-~ors i-tolst0 montior arel ticlo Se line monifto and ar adJ aent-to-the SO-blowidovines hii t hiilaflbn Cib .(IV-a .............

The monito's are used as. an aid in determining the source of SO blowdown radioactivity due to SO tube. rupture or a large prlmary-to-socondary leak.These-detectors-are-sa-fety-related-Class-l-B-and-meet-the-requtements-of-RO-,97-.

11.5,2.5.5 Main Steam Line High Bnergy Gamma (N-I k) Monitors:

Each main steam line is monitored by an ATL Nal scintillation detector, These detectors were installed to monitor the status of steaw generator primary to scoondary tube leaks and to provide a diagnostic tool for all individuals concerned with steam generator condition.

These detectors are designed to detect high energy gamma activity in the 6 to 7.2 MBV energy range.- High energy gamma activity int the main steam lines indicates the presence of N-16. The level of N-16 in the main steam lines is used to 11,5-44 Revision 14 RG2 STPEGS UFSAR Shielding for the SFP is adequate to protect plant personnel from exposure to radiation in excess of published guideline values as stated in Section 12.1. A minimum depth of approximately 13 ft of water over the top of an array of 193 (full core) assemblies with 42 hours4.861111e-4 days <br />0.0117 hours <br />6.944444e-5 weeks <br />1.5981e-5 months <br /> of decay is required to limit radiation from the assemblies to 2.5 mR/hr. or less.The FHB Ventilation Exhaust System is designed to limit the offsite dose in the event of a significant release of radioactivity from the fuel, as discussed in Sections 12.3.3,15.7.4, and 9.4.2. However, no credit for the FHB Ventilation Exhaust System is taken in the LOCA and Fuel Handling accident in Chapter 15.The FHB is designed to prevent missiles from contacting the fuel. A more detailed discussion on missile protection is given in Section 3.5.In addition, space is provided for storage of fuel during refueling inside the RCB for 64 fuel assemblies in four 4 x 4 modules having 16-in. center-to-center spacing (Figure 9.1.2-1A).

These modules are firmly bolted in the floor.9.1.2.2 Facilities Description.

The: FHB abuts the south side of the RCB and is adjacent to the west side of the MEAB of each unit. The locations of the two FHBs are shown in the station plot plan on Figure 1.2-3. For general arrangement drawings of the spent fuel storage facilities, refer to Figures 1.2-39 through 1.2-48 as listed in Table 1.2-1.The spent fuel storage facilities are designed for the underwater storage of spent fuel assemblies and control rods after their removal from the reactor vessel. The spent fuel is transferred to the FHB and handled and stored in the spent fuel pool underwater.

The fuel is stored to permit some decay, then transferred offsite. For a detailed discussion of spent fuel handling, see Section 9.1.4.The SFP is located in the northwest quadrant of each FHB. The floor of the pool is at El. 21 ft-Il in., with normal water level at El. 66 ft-6 inches. The top of a fuel assembly in a storage rack does not extend above the top of the storage rack which is El. 39 ft-10 in. maximum. The fuel assemblies are.loaded into the spent fuel racks through the top and are stored vertically.

9.1.2.3 Safety Evaluation.

Units 1 and 2 of the STPEGS are each provided with separate and independent fuel handling facilities.

Flood protection of each FHB is discussed in Section 3.4.1.A detailed discussion of missile protection is provided in Section 3.5.The applicable design codes and the various external loads and forces considered in the design of the Fl-B are discussed in Section 3.8.4. Details of the seismic design and testing are presented in Section 3.7.Design of this storage facility in accordance with GDC 62 and RG 1.13 ensures a safe condition under normal and postulated accident conditions.

The of the spent fuel storage racks is maintained less than or equal to 1.00, even if unborated water is used to fill the spent fuel storage pool, by both the designs of the fuel assemblies and the storage rack and the use of administrative procedures to control the placement of burned and fresh fuel and control rods.Under accident conditions, the Keff is maintained well below 0.95 assuming 2200 ppm borated water.The boron concentration of the water in the spent fuel pool is maintained at or above the minimum 9.1-4 Revision 16 REQUIREMENTS FOR RELIABLE.

SPENT FUEL POOL LEVEL INSTRUMENTATION AT OPERATING REACTOR SITES AND CONSTRUCTION PERMIT HOLDERS All licensees identified in Attachment I to this Order shall have a reliable indication of the water level in associated spent fuel storage pools capable of supporting identification of the following pool water level conditions by trained personnel:

(1) level that is adequate to support operation of the normal fuel pool cooling system, (2) level that is adequate to provide substantial radiation shielding for a person standing on the spent fuel pool operating deck, and (3) level where fuel remains covered and actions to implement make-up water addition should no longer be deferred.1. The spent fuel pool level instrumentation shall include the following design features: 1.1 Instruments:

The instrumentation shall consist of a permanent, fixed primary instrument channel and a backup instrument channel. The backup instrument channel may be fixed or portable.

Portable instruments shall have capabilities that enhance the ability of trained personnel to monitor spent fuel pool water level under conditions that restrict direct personnel access to the pool, such as partial structural damage, high radiation levels, or heat and humidity from a boiling pool.1.2 Arrangement:

The spent fuel pool level instrument channels shall be arranged in a manner that provides reasonable protection of the level indication function against missiles that may result from damage to the structure overthe spent fuel pool. This protection may be provided by locating the primary instrument channel and fixed portions of the backup instrument channel, if applicable, to maintain instrument channel separation within the spent fuel pool area, and to utilize inherent shielding from missiles provided by existing recesses and corners in the spent fuel pool structure.

1.3 Mounting

Installed instrument channel equipment within the spent fuel pool shall be mounted to retain its design configuration during and following the maximum seismic ground motion considered in the design of the spent fuel pool structure.

1.4 Qualification

The primary and backup instrument channels shall be reliable at temperature, humidity, and radiation levels consistent with the spent fuel pool water at saturation conditions for an extended period. This reliability shall be established through use of an augmented quality assurance process (e.g., a process similar to that applied to the site fire _p-6tection program).1.5 Independence:

The primary instrument channel shall be independent of the backup instrument channel.1.6 Power supplies:

Permanently installed instrumentation channels shall each be powered by a separate power supply. Permanently installed and portable instrumentation channels shall provide for power connections from sources independent of the plant ac and dc power distribution systems, such as portable generators or replaceable batteries.

Onsite generators used as an alternate power source and replaceable batteries used for instrument channel power shall have sufficient capacity to maintain the level indication function until offsite--'Attachment 2 resource availability is reasonably assured.1.7 Accuracy:

The instrument channels shall maintain their designed accuracy following a power interruption or change in power source without recalibration.

1.8 Testing

The instrument channel design shall provide for routine testing and calibration.

1.9 Display

Trained personnel shall be able to monitor the spent fuel pool water level from the control room, alternate shutdown panel, or other appropriate and accessible location.

The display shall provide on-demand or continuous Indication of spent fuel pool water level.2. The spent fuel pool instrumentation shall be maintained available and reliable through appropriate development and implementation of the following programs: 2.1 Training:

Personnel shall be trained in the use and the provision of alternate power to the primary and backup instrument channels.2.2 Procedures:

Procedures shall be established and maintained for the testing, calibration, and use of the primary and backup spent fuel pool instrument channels.A /2.3 Testing and Calibration:

Processes shall be established and maintained for.Lscheduling and implementing necessary testing and calibration of the primary and backup spent fuel pool level instrument channels to maintain the instrument channels at the design accuracy.

NEI 12-02 (Revision 1)...Mgi-st 2012 The three critical levels that must be monitored in a spent fuel pool are discussed below.It should be noted that continuous indication from a single instrument over the entire span from level 1 to level 3 is not required but could be used. If more than one instrument is used to monitor the entire span, that set of instruments constitutes a single channel satisfying either the primary or backup instrument channel requirement (refer to Figure 1 below).A visual representation of monitoring levels 1, 2 and 3 and the associated requirements for monitoring between the points are presented in Figure 1. The minimum requirements apply to the separation distance between level indications and support development of appropriate response procedures.

These requirements are separate from the instrument channel design accuracy discussed in section 3, which apply to either discrete or to continuous instruments.

Figure 1 2.3.1. Level 1 -level that is adequate to support operation of the normal fuel pool cooling system A typical fuel pool cooling system design includes a combination of weirs and/or vacuum breakers that prevent siphoning of the pool water level, below a minimum level, in the event of a piping rupture that can affect the SFP level.Level 1 represents the HIGHER of the following two points: 3 NEI 12-02 (Revision 1)August 2012 K0 The level at which reliable suction loss occurs due to uncovering of the coolant inlet pipe, weir or vacuum breaker (depending on the design), or The level at which the water height, assuming saturated conditions, above the centerline of the cooling pump suction provides the required net positive suction head specified by the pump manufacturer or engineering analysis.This level will vary from plant to plant and the instrument designer will need to consult plant-specific design information to determine the actual point that supports adequate cooling system performance.

2.3.2. Level

2 -level that is adequate to provide substantial radiation shielding for a person standing on the spent fuel pool operating deck Level 2 represents the range of water level where any necessary operations in the vicinity of the spent fuel pool can be completed without significant dose consequences from direct gamma radiation from the stored spent fuel. Level 2 is based on either of the following:

• 10 feet (+/- 1 foot) above the highest point of any fuel rack seated in the spent fuel pools, or" a designated level that provides adequate radiation shielding to maintain personnel radiological dose levels within acceptable limits while performing local operations in the vicinity of the pool. This level shall be based on either plant-specific or appropriate generic shielding calculations, considering the emergency conditions that may apply at the time and the scope of necessary local operations, including installation of portable SFP instrument channel components.

Additional guidance can be found in EPA-400 (Reference 4), USNRC Regulatory Guide 1.13 (Reference

5) and ANSI/ANS-57.2-1983 (Reference 6).Designation of this level should not be interpreted to imply that actions to initiate water make-up must be delayed'until SFP water levels have reached or are lower than this point.2.3.3. Level 3 -level where fuel remains covered and actions to implement make-up water addition should no longer be deferred.Level 3 corresponds nominally (i.e., +/- 1 foot) to the highest point of any fuel rack seated in the spent fuel pool. Level 3 is defined in this manner to provide the maximum range of information to operators, decision makers and emergency response personnel.

Designation of this level should not be interpreted to imply that actions to initiate water make-up must or should be delayed until this level is reached.4 Nuclear Oiperafeing Company SoutM T7-a.r P/ed Fiedc Cenerolng Stsfion PO, Baox229 Wdsworth, Texas 77"483 I 1 1.A February 28, 2013 NOC-AE-13002959 10 CFR 50.4 10 CFR 2.202 U. S. Nuclear Regulatory Commission Attention:

Document Control Desk Washington, DC 20555-0001 South Texas Project Units 1&2 Docket Nos, STN 50-498, STN 50-499 Overall Integrated Plan Regarding Commission Order Modifying Licenses with Regard to Reliable Spent Fuel Pool Instrumentation (Order Number EA-12-051)

References:

1. Letter, Eric Leeds to E. D. Halpin, "Issuance of Order to Modify Licenses with Regard to Requirements for Reliable Spent Fuel Pool Instrumentation," March 12, 2012 (EA-12-051)

2. NRC Interim Staff Guidance JLD-ISG-2012-03, Compliance with Order EA-12-051, Reliable Spent Fuel Pool Instrumentation, Revision 0, August 29, 2012 3, Letter D. W. Rencurrel to NRC, "initial Status Report in Response to March 12, 2012 Commission Order Modifying Licenses with Regard to Requirements for Reliable Spent Fuel Pool Instrumentation (Order Number EA-12-051)n, dated October 24, 2012.On March 12, 2012, the Nuclear Regulatory Commission (NRC) issued an order (Reference 1)to STP Nuclear Operating Company (STPNOC).

Reference 1 directs STP Nuclear Operating Company to provide a reliable Indication of the water level in associated spent fuel storage pools. Specific requirements are outlined in Attachment'2 of Reference 1.Reference I required submission of an overall integrated plan, including how compliance will be achieved.

The final Interim staff guidance (Reference

2) was issued August 29, 2012 providing licensees an acceptable approach for complying with the order. The purpose of this letter Is to provide the overall integrated plan, Including a description of how compliance will be achieved pursuant to Section IV, Condition C.1 .a, of Reference I in accordance with the guidance in Attachment 2 to Reference I and the guidance in Reference

2. See the Enclosure for STPNOC's response to the requested information.

There are no new commitments In this letter.33650640 NOC-AE-13002959 Page 2 of 3 If there are any questions regarding this letter, please contact Robyn Savage at (361) 972-7438.I declare under penalty of perjury that the foregoing Is true and correct.Executed on: _ , Dennis L. Koehl President and CEO/CNO

Enclosure:

South Texas Project (STP) Overall Integrated Plan for Implementation of Unit 1 &Unit 2 Spent Fuel Pool Level Instrumentation to Meet NRC Order EA-12-051 C.i 33650640 NOC-AE-13002959 Page 3 of 3 cc: (paper copy)(electronic copy)Regional Administrator, Region IV U. S. Nuclear Regulatory Commission 1600 East Lamar Boulevard:

Arlington, TX 76011-4511 Balwant K..Singal Senior Project Manager U.S. Nuclear Regulatory Commission One White. Flint North (MS 8 B1)11555 Rookville Pike Rockville, MD 20852 NRC Resident Inspector U. S. Nuclear Regulatory Commission P. O. Box 289, Mail Code: MN`16 Wadsworth, TX 77483 C. M. Canady City of Austin Electric Utility Department 721 Barton Springs Road Austin, TX 78704 U. S. Nuclear Regulatory Commission Director of Office of Nuclear Regulation One White Flint North (MS 13 H 16M)11555 Rockville Pike Rockvllle, MD 20852-2738 A. H. Gutterman, Esquire Morgan, Lewis & Bocklus LLP Balwant K. Singal U. S. Nuclear Regulatory Commission John Ragan Chris O'Hara Jim von Suskil NRG South Texas LP Kevin Polio Richard Pena City Public Service Peter Nemeth Crain Caton & James, P.C.C. Mele City of Austin Richard A. Ratliff Texas Department of State Health Services Alice Rogers Texas Department of State Health Services ('33650640 ENCLOSURE NOC-AE-13002959 South Texas Project (STP)Overall Integrated Plan for Implementation of Unit 1 & Unit 2 Spent Fuel Pool Level Instrumentation to Meet NRC Order EA-12-051 Page 1 of 12 I STP Overall Integrated Plan for Implementation of Unit 1 & Unit 2 Spent Fuel Pool Instrumentation to Meet NRC Order EA 12-051 Revision:

00 1.0 OVERALL INTEGRATED PLAN INTRODUCTION This document provides the overall Integrated Plan (the "Plan") which the STP Nuclear Operating Company ("STPNOC")

will implement for Units I and 2 to comply with the requirements of NRC Order EA-1 2-051, "Issuance of Order to Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation" (Ref.2), (the "ORDER"), NRC Interim Staff Guidance JLD-ISG-2012-003

[Rev.0] (Ref.3), (the "ISG"), and NEI Report 12-02[Rev.1] ("NEI 12-02").This Plan follows the format and provides all of the information on the STP 1 & 2 Integrated Plan that is required in NEI 12-02 [Rev.1] (Ref.1), Section A-2-2. Throughout this Plan, any reference to NEI 12-02 and the ISG will be based on the revisions above.Any reference to NEI 12-02 will include compliance to the clarifications and exceptions to NEI 12-02 required by the Interim Staff Guidance, Rev. 0.In response to the NRC requirements, STPNOC will provide-two channels of independent, permanently-installed, wide-range spent fuel pool level instrumentation

("SFPLI"), for the spent fuel pool ("SFP") of each unit. The spent fuel pool for each unit is independent and not interconnected in any way. For each SFP, the instrumentation provided for each channel will utilize the same technology, as permitted by the NEI 12-02 [Rev.1]. The spent fuel pool level instrumentation will provide continuous level indication for each SFP on both the Primary and Backup Channels.Both the Primary and Backup Channel/Instrument location and display of the SFP level will be independently mounted in each units Radwaste Control Room in the Mechanical Electrical Auxiliary Building (MEAB), which is an accessible post-event location.

Other locations are still being considered.

Both the Primary and Backup Channel remote, non-safety related indication of the SFP level will also be provided in each unit's Control Room via input to the Plant Computer.The instrumentation systems will not be safety-related, but will meet the requirements for augmented quality in accordance with NEI 12-02 [Rev.1] and the ISG as described below.Since all of the potential suppliers have not completed development, the information in this Plan is based on the overall strategy and on information which, based on current information from potential suppliers, is thought to envelope the systems being developed for this application.

If there are any changes to the requirements in NRC JLD-ISG-2012-003

[Rev.0] and NEI 12-02 [Rev. 1], relief from the requirements and schedule documented in this Plan may be required, in accordance with Section 12.0. Any required changes to this Plan will be defined in the periodic status reports submitted to the NRC.2.0 APPLICABILITY: ( Plan applies to the spent fuel pools for South Texas Project Unit I and Unit 2.Page 2 of 12 STP Overall Integrated Plan for Implementation of Unit I & Unit 2 Spent Fuel Pool Instrumentation to Meet NRC Order EA 12-051 Revision:

00 3.0 SCHEDULE: The installation of reliable spent fuel pool level instrumentation for the spent fuel pool associated with Unit .1 Is scheduled for completion prior to 10/28/2015, which is the end of the second refueling outage (1REI9) following submittal of this Plan.The installation of reliable spent fuel pool level instrumentation for the spent fuel pool associated with Unit 2 is scheduled for completion prior to 4/29/2015, which is the end of the second refueling outage (2RE17) following submittal of this Plan.Unit I Milestones are as follows: 0 Design/Engineering

-September of 2014 0 Purchase of instruments

& equipment-February of 2015 0 Receipt of equipment

-June of 2015* Unit 1 Installation

& Functional Testing -October of 2015 Unit 2 Milestones are as follows: a Design/Engineering

-December of 2013* Purchase of instruments

& equipment

-August of 2014* Receipt of equipment

-November of 2014* Installation

& Functional Testing -April of 2015 Consistent with the requirements of the ORDER and the guidance from NEI 12-02 [Rev. 1], status reports will be generated In six (6) month intervals from the submittal of this Plan.4.0 IDENTIFICATION OF SPENT FUEL POOL WATER LEVELS: The STP Unit 1 and 2 spent fuel pools are essentially identical.

The following SFP elevations are identified:

• The bottom of the pool Is at Plant El. 21 ft. 11 in.* The top of the SFP racks is approximately at Plant El. 39 ft. 10 in.* The minimum Limiting Condition for Operation SFP level is Plant El. 62 ft.0 Normal SFP water level is at Plant El. 66 ft. 6 in.0 Non-safety related level switch alarms are activated at Plant El. 67 ft. on high level and Plant El. 66 ft. on low level.* The spent fuel pool deck is at Plant El. 68 ft.The required key SFP water levels, per guidance of NEI 12-02 [Rev. 1] and ISG (with clarifications and exceptions), are as follows: Page 3 of 12 STP Overall Integrated Plan for Implementation of Unit I & Unit 2 Spent Fuel Pool Instrumentation to Meet NRC Order EA 12-051 Revision:

00 4.1 LEVEL 1: Level adequate to support operation of the normal fuel pool cooling system.LEVEL I represents the HIGHER of either the level at which reliable suction loss to the cooling pump occurs, or, the required NPSH (Nominal Pump Suction Head) of the cooling pump.Loss of reliable suction to SFP cooling pumps. For the purposes of this Plan, this level will conservatively be placed at Plant El. 64 ft. 2 in. This allows for just over I ft. of SFP water level above the top of the suction inlet flange (SFP Cooling Pump 14 in. suction line with centerline of suction inlet flange at Plant El. 62 ft. 6 in.)which will be sufficient for NPSH. (Ref. 9)Therefore, considering the top of SFP fuel storage rack is at Plant El. 39 ft. 10 in., the indicated level on either the Primary or Backup Instrument Channel of greater than 24 ft. 4 in. above the top of the SFP fuel storage racks based on the design accuracy for the instrument channel per NEI 12-02 [Rev.1], for both the Primary and Backup Instrument Channels, is adequate for normal SFP cooling system operation.

LEVEL I = Plant El. 64 ft. 2 in or 24 ft. 4 in. water level above the top of the SFP fuel storage rack 4.2 LEVEL 2: Level adequate to provide substantial radiationlshielding for a person standing on the SFP operating deck.Indicated level on either the Primary or Backup Instrument Channel of greater than 10 ft. above the top of SFP stored fuel assemblies based on current guidance in NRC RG 1.13 [Rev.2] (Ref. 4) will achieve substantial radiation shielding.

..Requirements on substantial SFP radiation shielding is also given in ANSI/ANS-57.2-1983 (Ref. 5), and states that radiation shall not exceed 2.5 mRem/hr, but the minimum water depth to achieve this Is undefined.

NRC RG 1.13 [Rev.21 took exception to using dose rates as design Input for minimum SFP water level, and instead defined the minimum level as 10 ft. above the stored fuel assemblies.

STPNOC elects to use the conservative approach of defining the top of the fuel rack as a basis for measurement.

Therefore, indicated level on either the Primary or Backup Instrument Channel of greater than 10 ft. above the top of the SFP fuel storage rack, based on the design accuracy of the instrument channel per NEI 12-02 [Rev.1], for both the Primary and Backup Instrument Channels, ensures there is adequate water level to provide substantial radiation shielding for a person standing on the spent fuel pool operating deck.Page 4 of 12 STP Overall Integrated Plan for Implementation of Unit 1 & Unit 2 Spent Fuel Pool Instrumentation to Meet NRC Order EA 12-051 Revision:

00 LEVEL 2 = Plant El 49 ft. 10 in. or 10 ft. water level above the top of the SFP fuel storage rack.4.3 LEVEL 3: Level where the fuel remains covered.As stated above, STPNOC elects to use the conservative approach of defining the top of the fuel rack as a basis for measurement.

The installation of the SFPLI sensorwill be such that it will measure as close as possible to the top of the SFP fuel rack. Indicated level on either the Primary or Backup Instrument Channel of greater than ' ft. above the top of SFP fuel storage racks based upon the design accuracy of the instrument channel per NEI 12-02 [Rev.1], for both the Primary and Backup Instrument Channels, satisfies the NEI 12-02 [Rev.2] requirement of E 1 ft.from the top of the fuel rack. This monitoring level ensures there is adequate water level above the stored fuel seated in the SFP fuel storage rack.LEVEL 3 = Plant El 40 ft. 4 In. or 6 In. water level above the top of the SFP fuel storage rack.5.0 INSTRUMENTS:

Both the Primary and Backup Instrument Channels will utilize permanently-installed instruments.

The design of the primary and backup instruments will be consistent with the requirements by NEI 12-02 [Rev.1], the ISG, and this Plan.The current plan is for both channels to utilize Guided Wave Radar, which functions according to the principle of Time Domain Reflectometry (TDR). A generated pulse of electromagnetic energy travels down the probe. Upon reaching the liquid surface the pulse is reflected and based upon reflection times level is inferred.

The measured range will be continuous from the high pool level elevation (67') to the top of the spent fuel racks (Ref. 8). However, STP is still evaluating other designs for this application.

Any changes to the current plan will be reported in the 6 month update letter.The supplier for the SFP instrumentation will be chosen by a competitive bidding process completed after submittal of this Plan, so the information in this Plan is based on the overall strategy and on available information from potential supplier's information on: systems being developed for this application.

5.1 Primary

(fixed) Instrument Channel The Primary Instrument Channel level sensing components will be located in the northeast corner of the Spent Fuel Pool, as shown in Attachment

1. The primary instrument channel will provide continuous level indication over a range from Plant El. 40 ft. 4 in. (LEVEL 3) to Plant El. 67 ft. (SFP high level alarm) or a range of 26 ft.8 in. In addition, the capability for discrete level indications at LEVEL1, LEVEL 2 and LEVEL 3, as described in Section 4.0, will be available.

Page 5 of 12 STPEGS UFSAR Shielding for the SFP is adequate to protect plant personnel from exposure to radiation in excess of published guideline values as stated in Section 12.1. A minimum depth of approximately 13 ft of water over the top of an array of 193 (full core) assemblies with 42 hours4.861111e-4 days <br />0.0117 hours <br />6.944444e-5 weeks <br />1.5981e-5 months <br /> of decay is required to limit radiation from the assemblies to 2.5 mR/hr. or less.The FHB Ventilation Exhaust System is designed to limit the offsite dose in the event of a significant release of radioactivity from-the fuel, as discussed in Sections 12.3.3, 15.7.4, and 9.4.2. However, no credit for the FHB Ventilation Exhaust System is taken in the LOCA and Fuel Handling accident in Chapter 15.The FHB is designed to prevent missiles from contacting the fuel. A more detailed discussion on missile protection is given in Section 3.5.In addition, space is provided for storage of fuel during refueling inside the RCB for 64 fuel assemblies in four 4 x 4 modules having 16-in. center-to-center spacing (Figure 9.1.2-1A).

These modules are firmly bolted in the floor.9.1.2.2 Facilities Description.

The FHB abuts the south side of the RCB and is adjacent to the west side of the MEAB of each unit. The locations of the two FHBs are shown in the station plot plan on Figure 1.2-3. For genefal arrangement drawings of the spent fuel storage facilities, refer to Figures 1.2-39 through 1.2-48 as listed in Table 1.2-1.The spent fuel storage facilities are designed for the underwater storage of spent fuel assemblies and control rods after their removal from the reactor vessel. The spent fuel is transferred to the FHB and handled and stored in the spent fuel pool underwater.

The fuel is stored to permit some decay, then transferred offsite. For a detailed discussion of spent fuel handling, see Section 9.1.4.The SFP is located in the northwest quadrant of each FHB. The floor of the pool is at El. 21 ft-1I in., with normal water level at El. 66 ft-6 inches. The top of a fuel assembly in a storage rack does not extend above the top of the storage rack which is El. 39 ft-10 in. maximum. The fuel assemblies, are loaded into the spent fuel racks through the top and are stored vertically.

9.1.2.3 Safety Evaluation.

Units 1 and 2 of the STPEGS are each provided with separate and independent fuel handling facilities.

Flood protection of each FHB is discussed in Section 3.4.1.A detailed discussion of missile protection is provided in Section 3.5.The applicable design codes and the various external loads and forces considered in the design of the FHB are discussed in Section 3.8.4. Details of the seismic design and testing are presented in Section 3.7.Design of this storage facility in accordance with GDC 62 and RG 1.13 ensures a safe condition under normal and postulated accident conditions.

The Keff of the spent-fuel storage racks is maintained less than or equal to 1.00, even if unborated water is used to fill the spent fuel storage pool, by both the designs of the fuel assemblies and the storage rack and the use of administrative procedures to control the placement of burned and fresh fuel and control rods.Under accident conditions, the KIf is maintained well below 0.95 assuming 2200 ppm borated water.The boron concentration of the water in the spent fuel pool is maintained at or above the minimum 9.1-4 Revision 16 cul ooPo4-RC-oo03

[ Excessive RCS Leakage Rev. 18 Page 53 of 127 Addendum 9 ~~~Basis.BssP f7 STEP DESCRIPTION FOR OPOP04-RC-0003 STEP 3.0 STEP: CHECK Trends For Any Of The Following Indications Of RCS Leakage:* Rad Monitor RT8011 Particulate

-Rising* Reactor Coolant Drain Tank Level -Rising* Pressurizer Relief Tank Level -Rising* RCB Normal Sump Level -Rising PURPOSE: To determine if leakage is from RCS and not CVCS.BASIS: Indication of RT801 1, RCDT, PRT or RCB Normal Sump levels rising will confirm that the leakage is from RCS and not CVCS which is normally tied to the RCS.ACTIONS: Monitor trends from RT801 1, RCDT, PRT or RCB Normal Sump.INSTRUMENTATION:

Level indications located on CP004 and various plant computer monitors located in control room. Radiation Monitor Computer RM- 11.CONTROL/EOUIPMENT:

N/A KNOWLEDGE:

N/A Th This Procedure is Applicable in Modes 1, 2, 3, and 4

.OPOPO03-ZG-0007 Rev. 71 Page 17 of 216 Plant Cooldown 3.57 Minimize the time at lowered RCS inventory (fuel in the reactor with level at or below the reactor vessel flange). Controls for Infrequently Performed Evolution per 0PGP03-ZA-0506, Tests or Evolutions Requiring Additional Controls, and OPGP03-ZO-0049, Conduct of Tests or Evolutions Requiring Additional Controls, SHALL be in place prior to lowering RCS level below 0% Pressurizer Cold Calibration Level elevation (elevation 52 ft 2 in) at Step 9.30.3.58 WHEN Steam Generator (SG) temperature is lowered, THEN SG narrow range level indication will indicate higher than actual level.3.59 Addendum 7 contains a list of conditions that should be met prior to taking credit for using the Steam Generators as a decay heat removal means while in Mode 5.3.60 During plant cooldown, all SGs will normally be connected to the steam header to assure a uniform cooldown of the RCS. (UFSAR 5.2.2.11.3) 3.61 The Main Steam lines upstream of the MSIVs may require periodic blowdown for moisture control. This can be accomplished by performing Addendum 13. MONITOR the following "MAIN STEAM OUTLET DRIP LEG LEVEL SWITCH" Plant Computer points-for-indi cations-of-moisture-buildupinrthe-Man-eStarn-Lines:

• LD7900, S/G 1A(2A) MS LN DRN FROM MS-2001* LD7901, S/G 1B(2B) MS LN DRN FROM MS-2002* LD7902, S/G 1C(2C) MS LN DRN FROM MS-2003* LD7903, S/G 1D(2D) MS LN DRN FROM MS-2004 3.62 Deaerator Storage Tank Level SHALL be maintained in normal band of 60% to 80% when condenser vacuum is established.

Going below 60% level may affect condenser vacuum.(Ref. 2.111)3.63 The principles of OPGP03-ZO-0042, Reactivity Management Program, are in effect at all times-during-Operati-onsin-thisyprocedure.

3.64 Shutdown margin SHALL be verified adequate based on the RCS boron concentration.

3.65 IF planned to place the RCS in MODE 5 with reactor coolant loops NOT filled or MODE 6 AND planned to swap the CVCS Bed Demineralizers in service during RCS in MODE 5 with reactor coolant loops NOT filled or MODE 6 THEN FLUSH the oncoming Demineralizers per OPOP02-CV-0004, Chemical and Volume Control System Subsystem PRIOR TO entering RCS in MODE 5 with reactor coolant loops NOT filled and MODE 6 conditions. (Ref 2.57)3.66 IF Personnel Air-Lock (PAL) doors are open in Mode 5, THEN Addendum 21, Closure of Personnel Air Lock Doors, is available to establish containment closure.This procedure, when completed, SHALL be retained.

OPOP03-ZG-0009 Rev. 59 Page 59 of 115 Mid-Loop Operation j Addendum 1 RCS/RHR Simplified Elevation Diagram Page 1 of I REACTOR COOLANT SYSTEM PRESSURIZER

-"7 RHR Cout)SECTION A-A HOT LEG STP D-0794 Rev 2 Qi OPOP03-ZG-0009 Rev. 59 Page'60 of 115 Mid-Loop Operation Ad dendum 2 RVWL Sensor Elevations I Page 1 of 1I NOTE* Top of Core is elevation 28 ft 2 inches.* SG spillover is elevation 34 ft 3.8 inches.SENSOR UPPER HEAD PLENUM UNCOVERED INDICATED INDICATED SENSOR LEVEL DESCRIPTION LEVEL (%) LEVEL (%)All Covered 100 100 46'4.75" Upper Head Full 1 64 100 45'3.4" Upper Head Partially Drained II -t 1 1 1 ii 2 0 100 39' 4.9" Plenum Full 3 0 85 34' 10.1" Plenum NOT Full (Enter Reduced Inventory) 4 0 66 33' 5.5" Top of Hot Leg Nozzle 5 0 48 32' 3" Hot Leg Centerline 6 0 33 31'0.5" Bottom of Hot Leg Nozzle 70 20 30' 1.6" Midway between Hot Leg Nozzle Sand Upper Core Plate 8 0 0 29' 2.7" Upper Core Plate CU2 OPOPo4-AE-o001 First Response To Loss Of Any Or Rev. 44 Page 29 of 7 All 13.8 KV Or 4.16 KV Bus Addendum.

4 Basis Basis Page 1 of 29 PROCEDURE PURPOSE This procedure provides guidelines for the initial response and stabilization of the plant in the event of a loss of any single or all 13.8 KV bus(es) or 4.16 KV Bus(es). This includes all 13.8 KV Auxiliary and Standby buses, and 4.16 KV buses with the exception of Buses 1K(2K), 1L(2L) and 1M. Loss of a 4.16 KV ESF bus is addressed as it indicates at least a partial loss of offsite and onsite AC power (ESF bus power can only be completely lost if both offsite and onsite power sources to the specific bus are lost).MAJOR ACTION CATEGORIES

• Provide interface with Emergency Operating Procedures and provide the instructions to establish the minimum equipment required to safely stabilize the unit.0 Identify actions associated with com ents to perforrnthefac.tion-withinta-specified-time-period-after the initiating event..DISCUSSION:

The electrical distribution system at STP has by design, a high degree of flexibility and ability to withstand casualties, especially the Class 1E alternating current systems. However throughout the nuclear industry Loss Of Offsite Power (LOOP) events have occurred as well as Station Blackout (loss of all offsite and onsite AC power) events.When dealing with a loss of offsite AC power, both complete and partial, with the Unit in Modes 1 or 2, the loss of an Auxiliary power bus will result in the loss of a Reactor Coolant Pump requiring a reactor trip because STP is not analyzed for operation with only three Reactor Coolant Pumps. In the event that no ESF bus is available the indication is that all offsite and onsite AC power has been lost requiring-t.tansition-to-the-Emergency-Operating-Proedurs.

Undrthese same condi-o-n-sSTP-has commint-d to shed the Channel I Load Sequencer from its power supply within the first 30 minutes after the initiating event, and if the associated battery bank has ajumpered cell then all the loads on DP 1201 and DP 1204 will be shed except for QDPS and SG PORVs.The initial response provided by this procedure is directed to the stabilization of critical plant parameters and then analyzes the extent of the loss of power. While this procedure does not identify the specific combination of buses that have been lost, it does identify the specific area of the power loss so that a procedure that is more specific to the method for power restoration can be referred to.This Procedure is Applicable in all Modes

....-0 First Response To Loss Of Any Or Rev. 44 Page 34 of 58 All 13.8 KV Or 4.16 KV Bus Addendum 4 Basis Basis Page 6 of 29." ........1.....STEP DESCRIPTION FOR OPOP04-AE-0001 STEP 3.0 STEP: CHECK 4.16 KV ESF Bus Status: a. ANY 4.16 KV ESF Bus NOT energized from offsite power (VERIFY the voltage on all three phases of each ESF Bus).b. VERIFY Applicable STBY DG(s) running c. VERIFY Applicable STBY DG(s) output breaker(s) closed to the associated 4.16 KV ESF bus PURPOSE: To determine the status of the 4.16 KV ESF buses and performs any corrective actions that can be performed under the current conditions.

BASIS--Thi-s step-at empts-to-start-SD G-for-a-de-energized-bus_-A-tso-this-step-ensures-output breaker is closed and if not determines the cause of the failure and provides steps to correct and energize the bus.If "4KV BUS O/C LOCKOUT" indicating lamp on applicable BSMP {CP003} is illuminated the bus cannot be energized until corrective maintenance is complete.ACTIONS: Determine if the SDG is available to be. started by checking for O/C lockout and other fault protection.

If available then perform the steps to start SDG and close output breaker.If the SDG is already running at this step, then determine the need to close the affected SDG breaker to energize the associated bus and close the breaker in the event that no faults exist. If a fault does exist, then the cause of the fault would have to be corrected before protective actuation device can be reset and the bus energized." INSTRUMENTATION:

N/A CONTROL/EQUIPMENT:

N/A KNOWLEDGE:

If the SDG has a 4.16 KV ESF Bus overcurrent lockout, SDG generator differential lockout or an SDG overspeed lockout then these faults will need to be corrected and reset to energize the bus.This Procedure is Applicable in all Modes O1 0POP04-AE-0004 Loss Of Power To One Or More 4.16 KV ESF Bus Addendum 14 Basis -Basis Page 1 of 18 PROCEDURE PURPOSE The purpose of this procedure is to restore power to any ESF bus which is not energized.

In the case where only one ESF bus is energized by a DG, and another one cannot be energized by the associated DG or offsite power, then steps are taken to operate breakers and disconnects to use the one running DG to supply key loads on another bus.MAJOR ACTION CATEGORIES

• Tie the operating DG to another bus via the emergency switchgear bus 1L(2L)." Energize at least one ESF bus from the Emergency Transformer.

• Control and load essential equipment on to the available ESF buses.-II DISCUSSION:

STP has committed under specific conditions related to loss of offsite and onsite power to energize at least two ESF buses from a running DG in order to energize specific loads needed to extend station battery life or provide availability of ESF equipment that is electrically powered from one of two specific ESF buses.-11 -11 This Procedure is Applicable in all Modes OPSPo3-EA-0002 Rev. 32 f Page 12of3 ESF Power Availability NOTE There are 5 possible lineups on Data Sheet 2, 3, and 4 for 13.8 KV XFMRS in the DESIGNATED Class 1E 4160 VAC Bus Power Source Table that meet Technical Specification requirements for being a power source for the 4.16 KV Buses:* (1) UAT supplying the AUX BUS and STBY BUS* (2) UAT supplying the AUX BUS and UNIT 1 STBY XFMR supplying the STBY BUS* (3) UNIT 1 STBY XFMR supplying the STBY BUS and the AUX BUS* (4) UAT supplying the AUX BUS and UNIT 2 STBY XFMR supplying the STBY BUS 0 (5) UNIT 2 STBY XFMR supplying the STBY BUS and the AUX BUS 5.2 COMPLETE Required ESF Power Train Data Sheet 2 through 4 by performing the following, steps.5-2 RECORD-aztuallbre-ake-r/di-scon-nect po sifio-fonr-sfhI---3 -KV-XFMRs, AUX BUS, STBY BUSES and from the 13.8 KV STBY BUS to the 480 VAC BUSES.* RECORD "CLOSED" breaker/disconnect positions by drawing a line at an angle through the breaker. ..* RECORD "OPEN" breaker/disconnect positions by drawing a CIRCLE around the breaker.EMER XFMR UNITAUXXFMR UNIT 1 STBY XFMR UNIT 2 STBY XFMR This example ,l illustrates:

R -(3) UNIT 1 STBY AUX BUS OPSP03-EA-0002 Rev. 32 Page 14 of 39 ESF Power Availability

6.0 Acceptance

Criteria NOTE Addendum 2, Two Physically Independent Circuits, provides a drawing of rights of way and offsite circuits to aide in the definition of "two physically independent circuits".

• Loss of one 13.8 KV Standby Bus to 4.16 KV ESF bus line constitutes loss of one required offsite source. (Reference 8.2)* Loss of two 13.8 KV Standby busses to 4.16 KV ESF bus lines constitutes loss of two required offsite sources. (Reference 8.2)* The preceding notes also apply when the 4.16 KV ESF bus is not energized by the 13.8 KV XFMR.* Step 6.1 applies during standby diesel inoperability.

• Step 6.2 applies during offsite independent circuits inoperability.

• Note and Precaution 3.28 should be referred to for additional clarification regarding allowable indication to be utilized when obtaining 345 KV switchyard voltage.6.1 Two physically independent circuits exist between the offsite transmission network and onsite Class 1E Distribution System as determined from Data Sheet 1, 2, 3, 4, and 9.(Technical Specifications 3.8.1. Lb, 3.8.1.1 .f, and 4.8.1.1. a)* North and South Bus in service with bus voltage:.o 340 KV" for NORMAL LINEUP OR o >356 KV for NORMAL LINEUP with UAT or Train B ESF LTC in "MANUAL" OR o _ 358 KV for all ALTERNATE LINEUPs OR voltage specified in the"Minimum-Vltage-flor-VariousAltemate-13..8-K-V--Bus-AI.ignments-!Addendum-

--of OPOP02-AE-0002, Transformer Normal Breaker and Switch Lineup.* Two of the following Rights of Way with a 345 KV line are available:

o NW Right of Way 1 (White Point 39)o NW Right of Way 2 (Elm Creek 27 OR WA Parish 39 OR Elm Creek 18)o Eastern Right of Way (Dow Velasco 27 OR Dow Velasco 18)* Two of the following 13.8 KV XFMRs are available:

o Unit Aux XFMR o Unit 1 Stby XFMR o Unit 2 Stby XFMR* Three 13.8 KV Standby Buses energizing the 4.16 KV ESF bus lines.

I I ' I I ~1 I-N ON IN"I IVfl INVAd 9NlIY~dO 0510H I dIarod SVX9I+/- U1 003 St c Cc cc~ C c C C t :ccc c c cc c c c c ~c -F-c CC CCC c , 0.cc c cc- C C C aCl c cc c cc-c~~~ Cr~' Ik -C+/-1 Q II ~ ~ t ~ .1 _-T IT~ C I CC T hAAFMV tKy7L _S4 RM-T -- 4 C,- CTT _ T _ T _0T 7 .I .L aY7/------ c r 4- -------[----I ~ 9 cc(+ c c c c c.cc cc ccc c CCC c cc c c ccc I5A~~*~~5of "dil cT0 ~ OLA4 4TTT4.*A~~f&gL A ~ ~ ~ ~ ~ ~ vý I, I -0T00000,, T.0. .0, -1 t "" ?~~a 0 CTI C_ Gl el l .q 0ooo0 0004000 0 -____E N___*~~~~~~~i~L 7N00440004*00 40 0 000004000000040~~~~~~~~--

0"00000004000004

-4#000u..4 400010400400010 0 t _ ____* u041z0 0~t 44 004"oOoo~~40

%~OtV 44 L CU3 TABLE 1.2 OPERATIONAL MODES.. ..MODE.1. PO 2. 51 3. HO 4. Kt.*5. CC 6. RE DWER OPERATION ARTUP)T STANDBY)T SHUTDOWN)LD SHUTDOWN EFUELING**

REACTIVITY CONDITION, Keff> 0.99> 0.99< 0.99< 0.99< 0.99< 0.95% RATED THERMAL POWER*> 5%< 5%0 0 0 0 AVERAGE COOLANT TEMPERATURE

> 350OF> 350 0 F> 350OF 350OF > Tavg> 200OF< 200OF< 140 0 F*Excluding decay heat..**FUel ihnthe reactor vessel tensioned or with the head with the vessel head closure bolts less removed.than fully SOUTH TEXAS -UNITS 1 & 2 1"9 CU4 (Page I of 283)O P Z 0 R.v* I ...O PGPO4-ZA-0307 I *I Preparation of Calculations Form 1 Calculation Cover Sheet CALCULATION COVER SHEET Page 1 Calculation No.: 13-DJ-006 Unit: 9 Bldg/Area/Sys:

VARIOUS Quality Class: A Priority Code: 2 Design Calculation iX ] Engineering Calculation Cog. Org.: ELECTRICAL Title: 125 VDC BATTERY FOUR HOUR COPING ANALYSIS Additional Dept:. N/A Signature:

Date: Review:. 'Additional Dept: N/A Signature:

Date: Review:.RPE Certification

[] Yes [X ] No Required: RPE.Signature:

N/A Date: _. _Registration No.: RPE Seal: "-]This calculation revision contains a change in the methodology as described in UFSAR Section Rev ORIGINAL CR actions tracking documents impacted by this revision to the calculation:

N/A (Page 14 of 283)(i: SOUTH TEXAS PROM CT ELECTRICAL CALCULATION SUBJECT 12s vbc BATTERY FOUR HOUR COPING ANALYSIS CALCULATION 13-DJ-006 REV. o 5.0 ACCEPTANCE CRITERIA 5.1 Battery Size The requiredbattery size, as calculated using the IEEE Standard 485-1978 (Ref, 6.3.2)methodology, must be less than or equal to the installed battery 'size, including the impact of minimum temperature and aging factors. This is determined by comparing the number of positive plates calculated to the actual number of positive plates for the installed battery.STP's defense-in-depth strategy requires the four Class IE DC channels to be "AC-Independent" for a minimum of four (4) hours, to facilitate copingwith a postulated loss of AC power event. The results of this calculation show the following:

5.1.]1. With no battery cellsjumpered out (i.e. 59 cell operationý, the Class IE DC Channel I can operate for a period of four (4) hours without battery charging sTupjport b

S-que Witii3-0Trwiiiit following the loss of Channel I battery- charging capability.

Class 1E DC Channel II can operate for a period of eight (8) hours without battery charging support and without shedding of any connected loads, Class IE DC Channels III and IV can operate for a period of four (4) hours without battery charging support and without shedding of any connected loads.2. With one battery cell jumpered out (i.e. 58 cell operation), the Class 1E DC Channel I can operate for a period of four (4) hours without battery charging support by manually de-energizing the ESF Load Sequencer "A" and shedding all but three loads on Panel 1201 within 30 minutes following the loss of Channel I battery charging capability.

Class 1E DC Channel II can operate for a period of eight (8) hours without battery charging support and without sheddingof any connected loads. Class IE DC Channel III can operate for a period of four (4)h ours-without-batter-y-char-ging-suppor.tand-w-ithoutshedding-ofany-connected loads. Class IE DC Channel IV can operate for a period of four (4) hours without battery charging support by manually shedding.

all but three loads on Panel 1204 within 30 minutes following the loss of Channel IV battery charging capability.

The loads on both 1201 and 1204 are breakers 13, 15, and 17 which are QDPS APC AI (C1) at 7 amps, QDPS APC A2 (C2) at 10 amps and Steam Generator IA PORV Servo Amplifier at 2 amps, according to EC-5008. This is a total of 19 amps which are then conVerted to power at 120 AC resulting in a power. of 2280* watts. The efficiency losses per EC-5008 are 2511 w. Summing these results in a power of 4791 W which are then converted back to DC amps at 125 VDC. .* resulting in the loads on the respective batteries EIV1201 and EIV 1204 of 38.328 amps after 30 minutes.Theminimum battery voltage that was used-in this calculation for all safety batteries to calculate the margin above was 106 volts. As an input to operations Emergency Operating Procedure OPOPO5-EO-ECO0

'Loss of All AC Power', the Page 14 of 283 (Page 15 of 283)(.SOUTH TEXAS PROJECT ELECTRICAL CALCULATION SUBJECT 125 VDC BA.TTERY FOUR HOUR COPING ANALYSIS C)LLCULATION 13-DJ-006 REV. 0 minimum bus voltage that any safety train can operate to is 105.5 VDC. Below 105.5 VDC it may be possible that some loads will have inadequate voltage to operate properly.

6.0 REFERENCES

6.1 Regulatory

6.1.1 South

Texas Project Technical Specifications and Bases, Amendment 198 for Unit I and Amendment 186 for Unit 2 6.1.2 South Texas Project Updated Final Safety Analysis Report (STP UFSAR)Chapter 8, Revision 16.6.1.3 Letter from T. H. Cloninger, STPEGS, to the NRC Document Control Desk, Revised Position of IOCFR50. 63, "Loss of All Alternating Current Power," dated......._ March 1,_19_9. (ST-HL-AE-5010)_

6.1.4. Letter

from T. H. Cloninger, STPEGS, to the NRC Document Control Desk, Supplemental Information to Revised Position of IOCFR50.63, "Loss of All Alternating Current Power," dated June 14, 1995 (ST-HLAE-5103)

6.1.5 Letter

from the Thomas W. Alexion, NRC, to'Mr. William T. Cottle, STPEGS, Revised Station Blackout (SBO) Position, South Texas Project, Units I and 2 (STP), dated July 24, 1995 (TAC Nos. M90061 and M90062) (ST-AE-HL-94257).

6.1.6 10CFR

50.59 Screen # 10-17753-5 Revision 0, "Revise Station Blackout Position to delete the need for a Coping Analysis" 6.2 Technical 6.2.1 Class lE 125 VDC Design Criteria Document, 4E520EQ0100, Rev 6 6.2.2 VTD-A363-002 Rev 6, "Instruction and Operating Manual 10 KVA Inverter" °6.2.3 VTD-A363-0045,Rev'l, "Vendor Technical Manual for Ametek Solidstate Controls 25KVA Inverter / Rectifier" 6.2.4 EC05036 Rev 8."DC Cable Sizing"'6.2.5 EC05037 Rev 5 "Maximum Allowable Length of AC Power Cables" 612.6 EC06038.Rev 9 "Power Cable Sizing Verification" a-6.2.7 VTD-S637-0009 Rev 1,.ESF.Load Settuencer for South Texas Project Electric Generating Station..6.2.8 DCN 9602493, dated 4/29/96 Page 15 of 283 REV. 23 OPOPO5-EO-ECOO LOSS OF ALL AC POWER PAGE 6 OF 7 ADDENDUM 4 VITAL DC BUS MONITORING STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED CAUTIONI I Do NOT allow battery voltages to drop to LESS THAN 105 VDC for plant equipment Iprotection.

NOTE Train A. B., and C bus voltages should be monitored for the duration of the event,.and their respective battery output breakers opened if bus voltages lowers to LESS THAN OR EQUAL TO 105.5 VDC in order to conserve the battery should a STBY DG become available.

4 MONITOR Class 1E 125 VDC system Train A, B, & C bus voltage.a. Train A AND B bus voltages a. PERFORM the following:

GREATER THAN 105.5 VDC 1) DISPATCH operator to perform ADDENDUM 3, FAILING AIR TO MSIVs AND MSIBs for all MSIV(s) and MSIB(s).2) WHEN ADDENDUM 3, FAILING AIR TO MSIVs AND MSIBs is complete, THEN GO TO Step 4.b of this Addendum.b. Train A, B, OR C bus voltages b. DISPATCH operator to open the GREATER THAN 105.5 VDC. associated battery output breaker: o "BTRY E1A11(E2All)

MAIN BKR" ElA11(E2A11)

BKR 1B (EAB 10')o "BTRY E1B1I(E2B11)

MAIN BKR" EIBII(E2B11)

BKR 1B (EAB 35')o "BTRY E1C11(E2ClI)

MAIN BKR" E1C11(E2CII)IBKR 1B (EAB 60')CHECK Sequencer(s) ready for RETURN TO procedure step in effect.restoration following bus energization CU5 SOUTH TEXAS PROJECT NUCLEAR OPERATING COMPANY D0527 STIE32563193 OPP5Z-O1Rev.

7 Page 1 ]of 23 I Emergency Communications I Quality Non Safety-Related Usag ffective Date: 12/03/09 j S. Korenek N/A N/A Emergency Response Division PREPARER TECHNICAL USER COGNIZANT ORGANIZATION Table of Contents Page 1.0 Purpose and Scope ....... ...........................................................................................................

2 2 .0 D efinition s ..........................................................................................................................................

2 3.0 R esponsibilities

.......................................................................................................

2 4.0 Emergency Communications System ...........................................................................................

3 5 .0 M ain ten an ce .....................................................................................................................................

15 6 .0 R eferen ces ........................................................................................................................................

15-7.0 -- Suiap i4 D i -entS-.Z..._.

...... ..........................................................................................................

15 Addendum 1, Communications Console Panel ..........................................................................

16 Addendum 2, Notification Methods to Offsite Agencies ...........................................................

17 Addendum 3, Station Public Address Selections

.......................................................................

18 Addendum 4, Related Maintenance Jacks ..................................................................................

19 Addendum 5, Portable Satellite Telephone

................................................................................

20 Addendum 6, Desktop Satellite Telephone Troubleshooting

.....................................................

21 OPGP05-ZV-0011 Rev. 7 Page 2 of 23 Emergency Communications

[1. Purpose and Scope 1.1 This procedure provides guidance in the use of emergency communications systems when responding to an emergency at the South Texas Project Electric Generating Station (STPEGS).2. Definitions 2.1 FTS 2001. System: A federal telephone system used by the Nuclear Regulatory Commission (NRC) and nuclear utilities for emergency communications.

2.2 RINGDOWN

LINE: A telephone line that does NOT require the operator or caller to dial a unuber to activate the circuit.2.3 UNIT OVERRIDE:

A circuit select switch (CSS) found on selected communications consoles, which when selected,.

activates prioritization circuitry for public address........-

announeementits-public address zones.3. Responsibilities 3.1 The Emergency Director, or designee, is responsible for activating the Emergency Notification System (ENS) to notify the NRC of a declared emergency, and to maintain communications with the NRC Operations Center.3.2 The Emergency Director, or designee, is responsible for activating the State/County ringdown line to notify State/County officials of a declared emergency.

3.3 The Radiological Manager or Radiological Director is responsible for activating the Health Physics Network (HPN) if requested by the NRC, to inform the Health Physics Section of physics information and response during a declared emergency at STP.3.4 The Manager, Information Technology or designee is responsible for the installation, testing, maintenance, and modifications of the emergency communications systems.

4. Emergency Communications System NOTE Refer to Addendum 2, Notification Methods to Offsite Agencies, for alternate telephone numbers and notification methods to be used throughout this procedure.

4.1 Emergency

Telephone Circuits 4.1.1 Emei:gency Notification System (ENS)The ENS is a telephone circuit provided by the NRC and is terminated on an FTS 2001 telephone.

The principal method of communications with the NRC is the ENS. The circuit may also be activated by the NRC. The ENS is activated to notify the NRC of declared emergency and to maintain communications with the NRC Operations Center.* IF the ENS is determined to be out of service and upon subsequent return to service, THEN notify the NRC Operations Center." ACTIVATE the ENS by lifting the handset on the telephone and dialing the appropriate number.4.1.2 State and County Ringdown Line* The State/County ringdown line is provided to notify State and County officials of a declared emergency.

The State/County ringdown line is an automatic ringdown telephone circuit terminated on a communications console OR an ORANGE telephone.

• ACTIVATE the State/County ringdown line by:-LIFTING the HANDSET on the ORANGE telephone OR-UTILIZING the communication console in accordance with Step 4.8, Coimnunications Console System.

OPGP05-ZV-0011 Rev. 7 Page 4of2l Emergency Communications

4.1.3 Health

Physics Network (HPN)The Health Physics Network (I-lN) is a telephone circuit provided by the NRC and is terminated on an FTS 2001 telephone.

It is to be used only at the request of the NRC. The HPN telephone is designed to provide communications with the NRC Health Physics Section and/or other nuclear power plants during a declared emergency.

STP health physics personnel MAY request a conference call with other nuclear power plants on the HiN by asking the NRC to connect the desired plant(s).* IF the HPN telephone line is determined to be out of service and upon subsequent return to service, THEN notify the NRC Operations Center.(IEN 89-19)0 ACTIVATE the HPN by lifting the handset on the telephone and diftling the appropriate number.4.1.4 STP Coordinator Ringdown Line* The STP Coordinator ringdown line is an automatic ringdown between the Qualified Scheduling Entity (QSE) and STP communications consoles." Utilize the communications console in accordance with Step 4.8, Communications Console System.4.2 Telephone System 4.2.1 The STP Telephone System consists of company owned and maintained telephone switching equipment and cable. The onsite system is connected to regular telephone services via an onsite demarcation point. The offsite services are provided by Verizon and Southwestern Bell Telephone.

Offsite commercial telephone services are augmented by a Center Point Energy owned and operated microwave system. The microwave system provides telephone and data services via tie lines into the Houston corporate offices. The corporate office telephone system interconnects into the local telephone system in Houston. The combined microwave and corporate office telephone systems provide augmentation to the normal local onsite -offsite telephone services at STP.

, ...... ...... =:=Emnlergen~cy Communications

.... .. ..:......4.2.2 Calling in (from offsite) may be accomplished in one of two ways:* Direct inward dialing (DID), OR* Calling the site number of (361) 972-3611 and using the automated attendant.

Direct inward dialing extensions begin with a 4, 7 or 8. All others must go through the automated attendant.

4.2.3 Calling

offsite (from onsite) may be accomplished in one of two ways:* DIAL 9-1-AREA CODE -telephone number, OR* DIAL 32-0 to Center Point Energy and have the Operator complete the call.4.2.4 Onsite calling is accomplished by dialing the desired extension number.S....... -42.-5-- -Two (-2)-mobile cellular-teleph ones are-provided to-Offsite-Field-Teams-as -a-- --back up to radio communications.

4.3 Portable

Satellite Telephone NOTE Portable, independent satellite telephones are provided to the Station as a backup to all company owned and commercial telephone equipment/services.

These telephones can be utilized for worldwide access via satellite.

4.3.1 Need clear view of the sky, outdoors, away from buildings and tall structures. -----T-ur-n-th e-ph on e-p ow-r-ono f-f2-P-ess-and-2 seconds see Addendum 5.4.3..3 Rotate and pull extend antenna into vertical position.4.3.4 To dial, press and hold the 0+ button until the display shows a + sign (the + sign is an international calling code), then proceed to dial just like any other long distance call (1 + area code + phone number).4.3.5 When you finish dialing, press OK to make the call.4.3.6 Talk with antenna above your head and vertical to the ground.4.3.7 When you complete the call press OK again to hang up.

, v. 7Page 6 of 23 Emergency Communications 4.3.8 Each portable satellite telephone is labeled with number and required codes for an outside caller to call back.4.3.9 To retrieve Voice Mail messages perform the following:

a Dial the satellite telephone number.* During the voice greeting, enter *a When prompted for your password, enter 1111.* Follow the voice prompts to: a. Play your messages.b. Record a Message.--~--~-- --c_--Change-your-greeting.-

___d. Access personal options.e. Make a call.4.4 Desktop Satellite Telephone NOTE Independent desktop satellite telephones are provided to the Station as a backup to all company owned and comnmercial telephone equipment/services.

These telephones can be utilized for worldwide access via satellite.

A desktop satellite telephone is maintained in both control rooms, both alport Centers, and the Emerg~eayQperations Facility.4.4.1 Although similar in many respects to a normal telephone, the desktop Satellite Telephone has some differences:

Pick up the telephone handset and listen; you should hear the normal steady state dial tone. The Satellite Terminal Box call status indicator should also shine green continuously.

A continuous orange indicator signifies acceptable but marginal signal strength.

The Satellite Terminal Box is located in the EOF Communications Room, and in the Unit 1 and 2 TSC Copy Room's on the communications rack.

OPGP05-ZV-011 Rev. 7 CPtge 7 of 23 Emiergenicy Communications1

a. If you hear nothing, there is possibly something wrong with your telephone or cable. Refer to Addendum 6, Desktop Satellite Telephone Troubleshooting.

b. If you hear a single tone interrupted every few seconds by silence, check that the signal strength indicator is orange or green. If not, there may be a problem with your SIM card. Refer to Addendum 6, Desktop Satellite Telephone Troubleshooting.

• . Dial 001 + area code + phone number. Once you have entered the phone number you will hear progress pips from the Iridiumn network. It can take up to 30 seconds for the Iridium network to connect a call, so a pause at this stage is not unusual.e Eventually you will hear the other end ringing, or hear a busy tone and voice message indicating why your call was unsuccessful.

When the other party answers the call status indicator will change from steady orge to flashing orange, indicating a call in progress., To terminate the call just hang up the handset. The call status light turns off.4.4.2 Each desktop satellite telephone is labeled with number and required codes for an outside caller to call back.4.4.3 To retrieveVoice Mail messages perform the following: " Dial the satellite telephone number." During the voice greeting, enter ** When prompted for your password, enter 11 11.L* Follow the voice prompts to: a. Play your messages.b. Record a Message.c. Change your greeting.d. Access personal options.e. Make a call.

OPGP05-ZV-0011 Rev. 7 Page 8 of 23 Emergency Communications

4.5 Radio

Communications 4.5.1 The Radio Communications System consists of repeaters, mobile, handheld, and base two-way FM transceivers licensed to STP Nuclear Operating Company by the Federal Communications Commission.

The radio repeaters are installed in a radio communications building at the base of the radio antenna tower onsite.The repeaters are supplied normal power from the plant power and emergency power from an automatic starting engine driven generator.

The generator is supplied fuel from a local fuel tank. The handheld, mobile and base stations are programmed to operate through the repeaters or direct.4.5.2 Radio communications with the Matagorda County Emergency Operations Center is accomplished by the use of a radio transmitter/receiver in the Security Central and. Secondary Alarm Stations, and a transmitter/receiver at the Matagorda County Sheriff s Office tuned to an STP radio frequency.

-5.5-3 ..O fflitf lTegfiil'r-di -commu_ fi-ions are accomw5lish-ed on SMTP Nu-ear Operating Company licensed radio channels.

The repeaters provide coverage-of the ten mile Emergency Planning Zone from one handheld radio to another handheld radio or to a base station.CAUTION IH-andheld radios SHALL NOT be used to transmit from inside the ESF Switch Gear Room, Control Room, Technical Support Center, Emergency Operations Facility, Auxiliary Shut Down Panel Rooms, Computer Rooms, nor within ten (10) feet of an open instrument cabinet, computer or computer terminals.

The only exceptions to the above restrictions are emergencies where a threat exists to the plant OR human safety and no other means of emergency communications are available.

4.5.4 PERFORM

the following to use a radio for communication:

• ALIGN the assigned radio channel on the handheld by selecting the appropriate channel number and Modes A and B for repeater or Mode C for direct communication." PRESS the microphone button and talk, keeping the microphone about 2 inches in front of the mouth, and* RELEASE the microphone button to receive, AND ADJUST the volume by turning the knob marked VOL.

OPGP05-ZV-0011 Rev. 7 Pag Emergency Communications

=1 = 771 I* ADJUST the squelch by turning the knob marked SQUELCH until noise is heard, then back until the speaker is quiet. This setting is for the maximum sensitivity, only on mobile radios." Communicate with other portable, mobile or base radio stations.4.6 Glenayre Paging System 4.6.1 The Glenayre Paging System is a tone system that may be activated from plant telephones or from an offsite touch-tone telephone.

The system has a range of over a 60-mile radius from the site. The system transmitters are connected to emergency power generators with automatic starting equipment.

4.6.2 Instructions

for activating the Glenayre Paging System are contained in OERPOI -ZV-1N03, Emergency Response Organization Notification.

4.7 Maintenance

Jack Communications System 4.7.1 A maintenance jack amplified and sound-powered telephone system is available for onsite communication between certain areas. Refer to Addendum 4, RelatedMaintenance Jacks. The system is powered by amplifiers on pre-designed circuits.

Each circuit may be activated or combined with another circuit by the proper selections on the system control panels located in each Control Room. The system has the capability to be voice activated.

The voice-activated circuit is one loop that interconnects each of the maintenance jack terminals into one circuit.4.7.2 IF it is desired to have amplified voice communications, TI-TEN PERFORM the following:

• SELECT the desired zones on the selection panel in the Control Room.-INSERT a headset plug into one of the jack stations marked 1 or 2 at the area.-INSERT a headset plug into the jack marked plant for voice-powered communications at the desired jack station.

OPGP05-ZV-0011 Re.7 Page 10 of 23 Emergency Communications

4.8 Communications

Console System 4.8.1 The communications console is an integrated communications panel and switching system which is subdivided into seven groups: direct line (ringdown), telephone, radio (RF), public address (PA), alarm system, conference, and voice direct line (VDL). Refer to Addendum 1, Communications Console Panel, for locations of the console controls.

Each communications group is composed of several two-position switches..

These positions are:* MONITOR -Top position (amber light will glow)* TALK/LISTEN

-Down position (green light will glow)4.8.2 When the TALK/LISTEN switch is activated (green light) for the State/County Ringdown line this locks out all other communications consoles.

When the call is completed, deactivate by depressing the TALK/LISTEN switch a second _--f i- -net-- et-ear thwe green light.4.8.3 These Consoles are installed in the Control Rooms, Auxiliary Shutdown Panel Rooms, Operations Support Ccntcrs, Technical Support Centers, Emergency Operations Facility, Security Force Supervisor's Office, Central and Secondary Alarm Stations, Simulator and in the Maintenance Office Facility.

During Refueling Outages, console(s) may be installed on the applicable units One Stop Shop.NOTE Many circuits may be monitored simultaneously.

These circuits are heard through the left ear if using the headset. The volume for the monitor position is controlled by the MONITOR VOLUME control located in the Handset/Headset Control Group.Usually the communicator operating the console will be talking (TALK/LISTEN switch is activated) on only one circuit at a time. These conversations will be heard through the right ear if using the headset. The volume control for the TALK/LISTEN position is controlled with the RECEIVE VOLUME control also located in the Handset/Headset Control Group.The communicator may actively communicate with all circuits simultaneously.

It is important to note that all circuits with the TALK/LISTEN switch activated will hear the communicators conversation, which may not be desirable.

To deactivate, depress the TALKJLISTEN switch a second time to clear the green light.

..... ... ... .........Emergency;P0 -z 0 i- Communications

....R -=-'" -P g l02 ..4.8.4 Direct Line (Ringdown)

Group Operation CAUTION Activating the circuit select switch (CSS) in the MONITOR (top position) will activate an Idle Circuit and cause the ringdown line to ring. The position switch SHALL be in the TALK/LISTEN (bottom position) before speaking.a. WHEN it is desired to place a call, THEN perform the following:

NOTE The next step will ringdown the other phone.C C-Activate-the-appropriate-circu:it--select -switc-h-in-the-T-ALKi/L-tS-T-EN--position.WHEN the phone is answered, THEN PRESS the push-to-talk button when speaking.WHEN communication is terminated, THEN DEACTIVATE the bottom TALK/LISTEN position switch.NOTE An audible signal will be heard through the speaker and the CSS red lamp will flash when a party is calling.b. WHEN a call is received, THEN perform the following:

• ACTIVATE the bottom TALK/LISTEN position switch.o WHEN it is desired to talk, THEN press the push-to-talk button when speaking.o WHEN communication is terminated THEN deactivate the bottom TALK/LISTEN position switch.

Emergency Communications

4.8.5 Telephone

Group Operation NOTE All normal site phone fuinctions are available through the console.a. IF it is desired to make a call, THEN PERFORM the following:

• ACTIVATE the circuit select switch for selected extension in the TALK/LISTEN (bottom) position AND WAIT until a dial tone is received on the headset or handset.* DIAL the number using the telephone keypad.* WHEN the number called answers, T14EN PRESS the push-to-talks-button-while-speaking....

• WHEN comm-nunication is terminated, THIEN DEACTIVATE the TALK/LISTEN switch.b. WHEN a call is received, THEN PERFORM the following:

NOTE An audible signal will be heard through the speaker and the CSS red light will flash when another party is calling.* ACTIVATE the circuit select switch (CSS) in the TALK/LISTEN (I tom position.* WHEN it is desired to talk, THEN PRESS the push-to-talk button while speaking.* WHEN communication is terminated THEN DEACTIVATE the two position TALK/LISTEN switch.* IF it is desired to place a call on hold, THEN ACTIVATE the MONITOR switch.

4PG.8.6...

..R Group Operation 23 Emergency Communications

4.8.6 Radio

Group Operation NOTE Radio channels may be monitored by moving the circuit select switch (CSS) to the MONITOR (top)position.a. IF it is desired to transmit a message on a radio frequency, THEN activate the circuit select switch to the TALK/LISTEN (bottom) position.b. PRESS the push-to-talk button when speaking.c. WHEN communication is terminated THEN deactivate the bottom TALK/LISTEN position switch.4.8.7 Plant Public Address and Alarm System NOTE Emergency alarm and public address override switch capabilities are found oi the communications console panels in the following locations:

all panels in each Unit's Control Room, and Technical Support Center, the Emergency Operations Facility, Central Alarm Station, Secondary Alarm Station, and the Simulator.

a. IF it is desired to make a public address announcement, TH-EN perform the following:

• SELECT the two position switch corresponding to the desired zone (listed on Addendum 3) that is to receive the announcement,* Activate the two position switch(es) to the TALK/LISTEN (bottom)position in the appropriate zone(s)." PRESS the push-to-talk button when speaking.* Deactivate the bottom TALK/LISTEN position switch at the-conclusion of the announcement.

Emergency Communications

b. Emergency Public Address Alarms and Announcement NOTE There are three public address emergency alarms: Assembly, Fire, and RCB Evacuation Alarm.Alarms will be broadcast as directed over the PA system. Alarm switches actuate for 8 seconds, then disconnect unless the PUSH-TO-TALK button on the handset is depressed.

• WHEN directed, THEN select the appropriate alarm.o WHEN the alarm is completed, THEN DEACTIVATE the alarm switch, activate the Unit override switch, AND make the appropriate emergency announcement over the PA system as directed.--WHEN-tlt-ahlh-i/a-fi-diffcemeJi -c-1 te-d,-= tE a-d _tvate all switches.4.8.8 Conference Network NOTE Loops may be monitored for informational purposes by selecting the MONITOR circuit select switch.a. PERFORM the following to establish group conference:

• VERIFY that all conferring parties are on the same loop.* VERIFY that all conferring parties on the loop have the circuit select switch (CSS) in the TALK/LISTEN (bottom) position.* WHEN it is desired to talk, THEN press the push-to-talkl button when speaking.* WHEN communication is terminated, THEN deactivate the bottom TALK/LISTEN position switch.

OPGP0-ZV-01 Rev. 7 ae1f 23 "~Emergency Communications

4.8.9 Voice

Direct Line (VDL)a. The Voice Direct Line (VDL) is a direct line from Quintron communication console to console.* Lift the handset on the appropriate console.* Activate the appropriate circuit selector switch on the communication to the TALK/LISTEN position.5. Maintenance

5.1 Information

Technology personnel SHALL maintain the emergency communications systems.5.2 Maintenance SHALL be done as required to keep the system in good operating condition and as committed to in license documents.

6. References 6.1 NUREG-0654/FEMA-REP-1, Criteria For the Development and Evaluation of Emergency Preparedness in Support of Nuclear Power Plants 6.2 South Texas Project Electric Generating Station Emergency Plan 6.3 OPGP07-ZA-00 11, Communications Systems 6.4 OERPOI -ZV-IN03, Emergency Response Organization Notification 6.5 IEN 89-19,. Health Physics Network 7. -Suppoe 7.1 7.2 7.3 7.4 7.5 7.6 trt.ec-u.mnen -s Addendum 1, Communications Console Panel Addendum 2, Notification Methods to Offsite Agencies Addendum 3, Station Public Address Selections Addendum 4, Related Maintenance Jacks Addendum 5, Portable Satellite Telephone Addendum 6, Desktop Satellite Telephone Troubleshooting

(

/".OOPGP05-ZV-0011 Rev. 7 Page 16 of 23 I-- -Emergency Communications

_Addendumd 1 Communications Console Panel Page 1 of _I PROGRAMMABLE FUNCTION. "F2" PROGRAMMABLE CIRCUIT (KEY)DESIGNATION STRIP PROGRAMMABLE FUNCTION 7F" FUNCTION "F3" LED BRIGHTNESS

.I 1 KEY POSITION AND CIRCUIT-STATUS---".

INDICATION LEDS CONTROL (U 0 oINCOMING SIG.o0 o 07 oVOLUME CON]SPEAKER 0 ]O C 0 o fn .11OC 0 ( 0.. .. .0 oO-a~o 0-] o0 Qo ob 0 0000 QJ] IVC MODE 0 0 0_________

VOLUME CP/DOWN 00cMF .IA Dmom 0000c 0 b 0 0T DIAL .PADE IP/DOWN)NAL rROL MUTE SELECT RECEIVE ONTROL SPEAKER MONITOR ACCE TWENTY FIVE CIRCUIT ACCESS KEYS (TALK/LISTEN-LOWER BUTTON)25 KEY DESKTOP OPGP05-ZV-0011 Rev. 7 Page 17 of 23 I .Emergency Communications I Addendum2 Notification Methodsto OffsiteAgencies 2Page Iof I ENS STATE (DPS, PIERCE) COUNTY (SHERIFF)

HPN 1-301-816-5100 1-979-541-4595 1-979-241-3205 1-301-816-51.00 1-301-951-0550 N/A 1-979-244-1178 1-301-951-0550 (ONLY when EOC is Activated)

NRC State/County ENS X Ringdown Line to the DPS, Disaster District Sub 2C (State of Texas) and the Matagorda County Sheriff's Office X (Matagorda County).Outside Telephone Lines. X X Satellite Telephone, X X Unit 1 Control Room Direct Line to Bay City. X X Microwave Line to Center Point Energy and call forwarded to the appropriate number.Rinadown Line to the STP Coordinator (OSE) and recuest y I I the call be forwarded to the appropriate number.Security Radio communication to the Matagorda County Sheriff's Office and request the call information be passed X X onto the appropriate number.

OPGP05-ZV-0011 I Rev. 7 Page 18 of 23 Emergency Communications Addendum 3 Station Public Address Selections Page 1 of I I I Unit 1 U it2 Units I & 2 UIhit 50 51 ALL ALL ALL Override Telephone Telephone-ZoneI I Zones Zones Zone4 Zone___________1

__ Zo re 2 1,2, & 3 .1,4 1one 34Zone 4 Electrical Auxiliary Building Electrical Au-xili~ry Building Unit I & 2 All Zones Essential Cooling Water Intake Structure Nuclear Support (EAB) (EAB) Yard simultaneously (ECWIS) Center (NSC), Mechanical Auxiliary Building Mechanical Aux liary Building .with activated Circulating Water Intake Structure Nuclear Training (MVlAB) (MAB) prioriti2,atiou (CWIS) Facility. (NTF)Isolation Valve Cubicle (IVC) Isolation Valve Cubicle (1VC) circui Lighting Diesel Generator Building (LD) Owner Reactor Containment Building Reactor Containment Building, Load Center Buildings 12J, 12K, 12L, Controlled Area (RCB) (RCB) 12M and the Electrical Load Center Fuel Handling Building (FHIB) Fuel Handling B iilding (FHB) Building (EL)Diesel Generator Building (DGB) Diesel Generator Building (DGB) H-ypochlorination Turbine Generator Building (TGB) Turbine Generatnr Building (TGB) Make Up Demineralizer (MUD)South/East Load Center Building Fire Pump House II North, East and West Gate Houses Units I and 2 Main and Standby Transformer Emergency Transformer Fuel Storage Building I Low Level Waste Building CWS Load Center-I Warehouse and Machine Shop_ __I Units I & 2 OPGP05-ZV-0011 Rev 7 Page 19 of 23[ .Em~ergen cy Communications E Addeindum 4 Related M aintenance Jacks Page I of1 UNIT 1 UNIT 2 TRANSFER SWITCH PANEL TRAIN A ESF1 ESF1 TRANSFER SWITCH PANEL TRAIN A ESF2 ESF2 TRANSFER SWITCH PANEL TRAIN B ESF8 ESF3 TRANSFER SWITCH PANEL TRAIN B ESF9 ESF9 TRANSFER SWITCH PANEL TRAIN C ESF10 ESF10 TRANSFER SWITCH PANEL TRAIN C ESF 1. ESFI I STANDBY DIESEL GENERATOR TRAIN A 1 SDG3 2SDG3 CONTROL PANEL STANDBY DIESEL GENERATOR TRAIN B 1 SDG2 2SDG2 CONTROL PANEL STANDBY DIESEL GENERATOR TRAIN C ISDGI 2SDGI_CQNTROL PANEL CHILLER CONTROL PANEL, TGI-17 TGI-17 COLUMN 18V BORIC ACID TANK ROOM RW-16 RW-16 ELE. 29' MAB, ROOM 076 CCW SURGE TANK ROOM MA-18 MA-18 ELE. 60' MAB., ESSENTIAL CHILLED WATER TRAIN A 1YD5 2YD8 INTAKE STRUCTURE ESSENTIAL CHILLED WATER TRAIN B IYD6 2YD9 INTAKE STRUCTURE ESSENTIAL CHILLED WATER TRAIN C 1YD7 2YD1O INTAKE STRUCTURE I I _I AUXILIARY FEEDWATER STORAGE TGI-12 TGI-12 TANK AREA, COLUMN 19Q I I OPGP05-ZV-0011 Rev. 7 Page 20 of 23;"ded .Emergency Communications I Addendum 5 Portable Satellite Telephone Page 1 of I Satellite Antenna Lock Release Button Signal Strength Indicator Status-Indicator....- Satellite Antenna SEarpiece Real-Time Clock Volume Control Keys-.....Battery- Charge-Indicator Display Scroll Bar Display and -,"-Status Indicators Message Key* Press to dial Press to Hang up Battery and Sim I-Alphanumeric

_Keypad Quick Access Key.-Headset Jack Power on/off key -Power Connector

---

Microphone Data Connector

-.."

OPGPo5-ZV-0011 Rev. 7 Page 21 of 23 Emergency Communications EL ddendurrn 6 ]. Desktop Sa~t.ellite Telephone Troubleshooting e_ Iof3 REMOTE SATELLITE TERMINAL (RST-100)Front Panel Status LEDS are located on the front panel, they show the RST-100 status.1. Power 2. Voicemail waiting 3. SMS waiting 4. Call status 5. Signal strength Once powered up, the RST-100 attempts to register with the Iridium network. The signal level LED uses color to indicate how strong the Iridium signal is at your location.Indicator color Signal strength tireen Orange Red 6trong Acceptable No signal, problem with installation In most cases the indicator will show green after a short period of approximately 15 seconds -orange indicates an acceptable but marginal signal strength.

If.the indicator remains red, there is a problem with the installation.

Rear Panel OPGP05-ZV-0011 Rev-. 7 Page 22 of 23 Emergency Communications This table provides information to help you troubleshoot problems encountered while using the Desktop Satellite Telephone (RST-l 00). If the problem continues contact IT Communications at extension 7000 or if the ERO/Storm Crew is activated the Communications Systems Supervisor.

QUESTION ANSWER No lights on the front panel of the Satellite Check power is connected.

Terminal Box.RST-100 fails to register with the Iridium service Press reset button located on the rear of the Satellite after 30 seconds. Terminal Box.-No dial tone. -C--Check if a-dala-call-is -in-progress-an d-power-i s-..connected and equipment is in a normal state'Cannot make call, two tone signal heard. Phone requires a PIN. See step 1 below.You can't make calls. Check that the antenna is properly mounted.Did you enter the number in international format?Check the signal strength.

If the signal is weak, wait a few minutes for thick cloud cover to move.Has a new SIM card been inserted?Vcm ri~n't r~.ceAve~

c,~ll~Clit-.rle flip nnti-.nnn TQ ;t nrnnp.rlv rriminfp.0 ou can't receive calls Check the signal strength.

If the signal is weak, wait a few minutes for thick cloud cover to move.Check the telephone ringer setting to see if it is off.The Voicemail indicator keeps flashing..

There is not enough memory available to store another message. Or there is a message waiting.Delete messages and free up some space.

OPGP05-ZV-0011 Rev. 7 Page 23 of 23 A e mEmergency Communications F Addendu Desktop Satellite Telephone Troubleshooting

/ Page 3 of 3 1. Personal Identification Number (PIN): Your RST- 100 may require a PIN, this will be indicated by the Signal light flashing Red and a distinctive dial tone consisting of two alternating tones.If a distinctive, two-tone dial tone is heard, one of two access codes is required -the SIM PIN or the PIN Unlock. Code (PUK).The SIM PIN is required if the two tones are of equal length. If so simply enter the four digit PIN (1 I1) and await a change of tone (up to ten seconds), then hang up. If the PIN was correct the phone will register and you. may proceed with normal use as described below.--.The- PUK-is required-if-the-PIN-has-been -ineorrectly-entered-three-times-and-is indicated when-the high tone is longer than the low tone. Contact communications at extension 7000 or if the ERO/Storm Crew is activated the Communications Systems Supervisor

2. Voicemail* If a Voicemail message has been left for you, the RST-100 flashes the Voicemail Waiting indicator.

The indicator is cleared whenever the user connects to the Voicemail retrieval number.3. Facsimile Support* The Iridium network does not support facsimile transmission.

CAI NOTE Step 3.0 will determine if leakage is actual RCS leakage.3.0 CHECK Trends For Any Of The Following Indications Of RCS Leakage:* Rad Monitor RT8 011 Particulate

-Rising* Reactor Coolant Drain Tank Level -Rising* Pressurizer Relief Tank Level -Rising* RCB Normal Sump Level -Rising.Go TO Step 5.0.4.0 PERFORM One Of The Following To Determine The RCS Leak Rate:* OPSP03-RC-0006, Reactor Coolant Inventory OR" DETERMINE the RCS leak rate using pressurizer level, VCT level, and comparing charging and letdown flows------------------------------------------------------

4

-OPOP04-RC-0003 Excessive RCS Leakage Rev. 18 Page 53 of 127 Addendum 9 _ Basis Basis Pa e 5 of77 STEP DESCRIPTION FOR OPOP04-RC-0003 STEP 3.0 STEP: CHECK Trends For Any Of The Following Indications Of RCS Leakage:* Rad Monitor RT801 1 Particulate

-Rising* Reactor Coolant Drain Tank Level -Rising* Pressurizer Relief Tank Level -Rising* RCB Normal Sump Level -Rising PURPOSE: To determine if leakage is from RCS and not CVCS.BASIS: Indication of RT801 1, RCDT, PRT or RCB Normal Sump levels ris ng will contirm that the leakage is from RCS and not CVCS which is normally tied to the RCS.ACTIONS: Monitor trends from RT801 1, RCDT, PRT or RCB Normal Sump.INSTRUMENTATION:

Level indications located on CP004 and various plant computer monitors located in control room. Radiation Monitor Computer RM-11.CONTROL/EOUIPMENT:

N/A KNOWLEDGE:

N/Aý I

-,OPOP02-RHI-0001 IRev. 63 Page 39 of 253 Residual Heat Removal System Operation (7.43 MONITOR. Plant Computer group RH-12 (8412) OR TREND the following points for the applicable pump: "RHR PUMP 1A(2A)""RHR PUMP IB(2B)""RHR PUMP 1C(2C)" RHFE0867 RHIA08 80 RI-FE0868 RH1A0881 RHFE0869 RHIA0882 CAUTION U -.. ."'-. ----- ---.. ---U D0 NOT start an RHK pump with vessel level below 32 ft 9 inch. (6 inches above hot leg centerline)

SWHIEN the DG is being paralleled ORR operated in parallel with offsite power, THEN the associated Trains "RHR PUMP" SHALL NOT be started or operated: (CR 05-4915)/.7.44 ENSURE the associated train's Emergency Diesel Generator for the pump to be started in the next step is NOT being paralleled OR operated in parallel with offsite power. (CR 05-4915).7.45 START the desired RHR pump: "RHR PUMP 1A(2A)"* ."RHR PUMP 1B(2B)."* "RHR PUMP 1 C(2C)" ('

0POP02-RH-0001 Rev. 63 Page 39 of 253 Residual Heat Removal System Operation 7.43 MONITOR. Plant Computer group RH-12 (8412) OR TREND the following points for the applicable pump: "RHR PUMP lA(2A)""RHR PUMP 1B(2B)""RHR PUMP 1C(2C)" RHFE0867 RHIA08 80 RHFE0868 RHIA0881 RHIFE0 869 RHIA0882 CAUTION* DO NOT start an RHR pump with vessel level below 32 ft 9 inch. (6 inches above hot leg centerline)

• WHEN the DG is being paralleled OR operated in parallel with offsite power, THEN the associated Trains "RHR PUMP" SHALL NOT be started or operated: (CR 05-4915)1 7.44 ENSURE the associated train's Emergency Diesel Generator for the pump to be started in the next step is NOT being paralleled OR operated in parallel with offsite power. (CR 05-4915).7.45 START the desired RHR pump: "RHR PUMP 1A(2A)"* ."RHR PUMP IB(2B)""RHR PUMP 1C(2C)".(

CA2 OPOP04-AE-0001 First Response To Loss Of Any Or All 13.8 KV Or 4.16 KV Bus~Addendum 4 Basis Basis Page 1 of 29 ,j PROCEDURE PURPOSE This procedure provides guidelines for the initial response and stabilization of the plant in the event of a loss of any single or all 13.8 KV bus(es) or 4.16 KV Bus(es). This includes all 13.8 KV Auxiliary and Standby buses, and 4.16 KV buses with the exception of Buses 1K(2K), 1L(2L) and IM. Loss of a 4.16 KV ESF bus is addressed as it indicates at least a partial loss of offsite and onsite AC power (ESF bus power can only be completely lost if both offsite and onsite power sources to the specific bus are lost).MAJOR ACTION CATEGORIES

• Provide interface with Emergency Operating Procedures and provide the instructions to establish the minimum equipment required to safely stabilize the unit.___. Identify-actions-associated-with-conm-itments-to-per form-the-act-ion-within-a-spe-ified4ime-petiod-after the initiating event.DISCUSSION:

The electrical distribution system at STP has by design, a high degree of flexibility and ability to withstand casualties, especially the Class 1 E alternating current systems. However throughout the nuclear industry Loss Of Offsite Power (LOOP) events have occurred as well as Station Blackout (loss of all offsite and onsite AC power) events.When dealing with a loss of offsite AC power, both complete and partial, with the Unit in Modes 1 or 2, the loss of an Auxiliary power bus will result in the loss of a Reactor Coolant Pump requiring a reactor trip because STP is not analyzed for operation with only three Reactor Coolant Pumps. In the event that no ESF bus is available the indication is thatall offsite and onsite AC power has been lost requiring transilion to th-TEmergency Operating Procedures.

Under these same conditions STP has committed to shed the Channel I Load Sequencer from itspower supiSly within the first 30 minutes after the initiating event, and if the associated battery bank has a jumpered cell then all the loads on DP 1201 and DP 1204 will be shed except for QDPS and SG PORVs.The initial response provided by this procedure is directed to the stabilization of critical plant parameters and then analyzes the extent of the loss of power. While this procedure does not identify the specific combination of buses that have been lost, it does identify the specific area of the power loss so that a procedure that is more specific to the method for power restoration can be referred to.This Procedure is Applicable in all Modes Ii ~

OPOP04-AE-0001 First Response To Loss Of Any Or Rev. 44 Page 34 of 58 All 13.8 KV Or 4.16 KV Bus Addendum 4 Basis Basis Page 6 of 29 STEP DESCRIPTION FOR OPOP04-AE-0001 STEP 3.0 STEP: CHECK 4.16 KV ESF Bus Status: a. ANY 4.16 KV ESF Bus NOT energized from offsite power (VERIFY the voltage on all three phases of each ESF Bus).b. VERIFY Applicable.STBY DG(s) running c. VERIFY Applicable STBY DG(s) output breaker(s) closed to the associated 4.16 KV ESF bus PURPOSE: To determine the status of the 4.16 KV ESF buses and performs any corrective actions that can be performed under the current conditions.

BASISYThistep--ate-mpt-to-s-t-WSD-G-fo-a-de--ergi--d-bus-.

Also0tllistep ensures output breaker is closed and if not determines the cause of the failure and provides steps to correct and energize the bus.If "4KV BUS O/C LOCKOUT" indicating lamp on applicable BSMP {CP003} is illuminated the bus cannot be energized until corrective maintenance is complete.ACTIONS: Determine if the SDG is available to be started by checking for O/C lockout and other fault protection.

If available then perform the steps to start SDG and close output breaker.If the SDG is already running at this step, then determine the need to close the affected SDG breaker to energize the associated bus and close the breaker in the event that no faults exist. If_oal xkt-Ah-nAhe musgofhefaub ild aye tobcnm tedtd~heopratetiv actuation device can be reset and the bus energized.

INSTRUMENTATION:

N/A CONTROL/EQUIPMENT:

N/A KNOWLEDGE:

If the SDG has a 4.16 KV ESF Bus overcurrent lockout, SDG generator differential.

lockout or an SDG overspeed lockout then these faults will need to be corrected and reset to energize the bus.This Procedure is Applicable in all Modes OPOP04-AE-0004 Loss Of Power To One Or More 4.16 KV ESF Bus I ...........

LI ~Addendum 14 Basis. Basis Page 1 of 18 PROCEDURE PURPOSE The purpose of this procedure is to restore power to any ESF bus which is not energized.

In the case where only one ESF bus is energized by a DG, and another one cannot be energized by the associated DG or offsite power, then steps are taken to operate breakers*

and disconnects to use the one running DG to supply key loads on another bus.MAJOR ACTION CATEGORIES

• Tie the operating DG to another bus via the emergency switchgear bus IL(2L).* Energize at least one ESF bus from the Emergency Transformer.

• Control and load essential equipment on to the available ESF buses.DISCUSSION:

STP has committed under specific conditions related to loss of offsite and onsite power to energize at least two ESF buses from a running DG in order to energize specific loads needed to extend station battery life or provide availability of ESF equipment that is electrically powered from one of two specific.ESF buses.-.11 This Procedure is Applicable in all Modes OPSP03-EA-0002 Rev. 32 Page 12 of 39 ( ESF Power Availability NOTE There are 5 possible lineups on Data Sheet 2, 3, and 4 for 13.8 KV XFMRS in the DESIGNATED Class 1E 4160 VAC Bus Power Source Table that meet Technical Specification requirements for being a power source for the 4.16 KV Buses:* (1) UAT supplying the AUX BUS and STBY BUS* (2) UAT supplying the AUX BUS and UNIT 1 STBY XFMR supplying the STBY BUS o (3) UNIT 1 STBY XFMR supplying the STBY BUS and the AUX BUS* (4) UAT supplying the AUX BUS and UNIT 2 STBY XFMR supplying the STBY BUS* (5) UNIT 2 STBY XFMR supplying the STBY BUS and the AUX BUS 5.2 COMPLETE Required ESF Power Train Data Sheet 2 through 4 by performing the following steps.-5+2.-1 AUX BUS, STBY BUSES and from the 13.8 KV STBY BUS to the 480 VAC BUSES.-* RECORD "CLOSED" breaker/disconnect positions by drawing a line at an angle through the breaker. _ -RECORD "OPEN" breaker/disconnect positions by drawing a CIRCLE around the breaker..EMER XFMR UNIT AUX XFMR UNIT 1 STBY XFMR UNIT 2 STEY XFMR This example illustrates:

(3) UNIT I STBY.............

__ _ __ _XFMR suppj.ying h STBY BUS and the.. ....-....AUX BUS!......_ .......

OPSP03-EA-0002 7Rev. 32 Page 14 of 39 1ESF Power Availability

6.0 Acceptance

Criteria NOTE" Addendum 2, Two Physically Independent Circuits, provides a drawing of rights of way and offsite circuits to aide in the definition of "two physically independent circuits".

• Loss of one 13.8 KV Standby Bus to 4.16 KV ESF bus line constitutes loss of one required offsite source. (Reference 8.2)* Loss of two 13.8 KV Standby busses to 4.16 KV ESF bus lines constitutes loss of two required offsite sources. (Reference 8.2)* The preceding notes also apply when the 4.16 KV ESF bus is not energized by the 13.8 KV XFMR.* Step 6.1 applies during standby diesel inoperability.

• Step 6.2 applies during offsite independent circuits inoperability.

• Note and Precaution 3.28 should be referred to for additional clarification regarding allowable indication to be utilized when obtaining 345 KV switchyard voltage.6.1 Two physically independent circuits exist between the offsite transmission network and onsite Class 1E Distribution System as determined from Data Sheet 1, 2, 3, 4, and 9.(Technical Specifications 3.8.1.1.b, 3.8.1.1.f, and 4.8.1.1.1.a

)* North and South Bus in service with bus voltage: o >340 KV" for NORMAL LINEUP OR o 3 356 KV for NORMAL LINEUP with UAT or Train B ESF LTC in "MANUAL" OR o 358 KV for all ALTERNATE LINEUPs OR voltage specified in the"Minimum Voltage for Various Alternate 13.8 KV Bus Alignments" Addendum of OPOP02-AE-0002, Transformer Normal Breaker and Switch Lineup.* Two of the following Rights of Way with a 345 KV line are available:

o NW Right of Way 1 (White Point 39)o NW Right of Way 2 (Elm Creek 27 OR WA Parish 39 OR Elm Creek 18)o. Eastern Right of Way (Dow Velasco 27 OR Dow Velasco 18)Two of the following 13.8 KV XFMRs are available:

o Unit Aux XFMR o Unit 1 Stby XFMR o Unit 2 Stby XFMR Three 13.8 KV Standby Buses energizing the 4.16 KV ESF bus lines.

r-* I *

• I.vilya LOS~ bi;imi o 031VU 130 JVIS WOM100 C~ cc'~C C C C T -I --L: -T T ___T_ -TT-TZ T TJ I_-j CA3 TABLE 1. 2 OPERATIONAL MODES..,MODE 1. POWER OPERATION 2. STARTUP 3. HOT STANDBY 4. HOT SHUTDOWN REACTIVITY CONDITION, Keff> 0.99> 0.99< 0..99< 0.99< 0.99< 0.95% RATED THERMAL POWER*> 58X< 5%0 0 0 0 AVERAGE COOLANT TEMPERATURE

> 350OF> 350OF> 350OF 350OF > T> 200OF avg< 200OF< 140OF.5 6.COLD SHUTDOWN REFUELING**

• Excluding decay heat,.**Fuel in the reactor vessel with the tensioned or with the head removed.vessel head closure bolts less than fully SOUTH TEXAS -UNITS 1 & 2 CA6

> ...STPEGS UFSAR 3A4 WATER LEVEL (FLOOD) DESIGN The methods of analyais used to determine the design basis flood are discussed in Section 2.4, These methods are consistent with the requirements.

ofRegulatory Guide (RG) 1.59.The protection measures used to accommodate static arid dynamic flood loads on Category I structures generally fall under the category of "Incorporated barriets" as specified in regulatory positibn C. I of RG' 1.102..3.4.1 Flood Protection 3,4.1.1 &teral Flood Proteotjon Measures for Ses*ic Oatugory I Stmutures, The flooding due to a postulated Main Cooling Reservoir (MCR) embankment breach produces the maximum water level around the power block straotires as well as the oontr6lling water elevations fbr buoyancy calculations.

This. is also the controlling phenomena in deteraiilning the maximum water level at the Esseantial Cooling Water Intake Sttucture

-(WISý. Studies and analyses on the MCR embanledient have demonstrated that an adequate margin of safety'oan be maintained for all credible failure mehmaalams (Section 2,5.6). Accordingly, mechanistic effects (such as scour and.-- eesion-)-asaoakted.-w4.th.

a-postulated-

--.-- --The maximum water level on a vertical face at the south end of the plant structures is El.. 50,8 ft mean sea level N4SLJ5 which is El. 22.8 ft above plant grade., This maximum elevation occurs during a quasi-stoady-#tate condition after a breach of the MCR embanulient and is based on an instantaneous removal of alpproximatoly 2,000 ft of the embankent opposite the power block sructures, This maximum elevation occurs on. the. souih f~ce of the Fuel-Handlhing Building .(FHI) of Unit 1. The selection of postulated embankment breach widths and the, asvmptions mnade in determining the maximum flood elevations, are described in Section 2A.4.4.Total intidation of the Essential' Cooling Pond (ECP) occurs only under the conditioon of MCR embankment breach and 'does not affect the safe shutdown capability of th6 plant, The maximum water level ohlculated'to oocurat the BCWIS is El. 40.8 ft.Safey.4d1ated stniutdifes, q~es af~ind, comxpotifiets Nistd 1&Thble 3Z2A-K'Iare protectd-aginsft_&

-effects of external flooding by: I. Being designed to withstand the maximum flood level and aisooiated effects and temain functional (such as seismic Category I -structures and the Category I auxiliary feedwater-storage-tank):or

' ;2. ' Being housed .within seismic Category I structures which are designed as in item 1, above.Flood protectlon of safety-related structures, systems, and components is provided for postulated

'I flood levels and conditions described fn Section 2.4..4 Qi., Seismic Category I structures are designed to withstand the maxtimum flood le~'els by: ].3.4-1.Revision 13.

,(.]: STFBGS MFSAR 1. Having external walls and :slabs of structures designed to resist the hydrostatic and hydrodynamic foroes associated with surge-wave runup and steady-state water level, 2. Ensuring the overall, stability -of thb total structure against overturning and sliding due to the hydrostatic and hydrodynamic forces associated with surge-wave runup and steady state water level, and 3. Enisuring that the total structure will not float due to buoyancy forces.Figure 3.4-1 shows. a general section through the plant. Figure 3-A-2 shows the seismic. Category I Building xnxlmum steady-state water surface profile, and the corresponding relationship of sill elevations for entrances to seismic Category I buildings.

Table 3A4-1 shows the results of hydraulic loading and buoyancy calculations which were done for the various safety-related facilities, 'The water depths shown on this table were developed from the maximum water surface elevations presented in Table 2AA-3.An investigation of seismic Category I struotures has been made for the flood levels and-associated.

factors greater than 1.1. All exterior seismic Category I building openings are. located above the maximur steady-state flood level or are equipped with watertight doors when located below this" profile, except as' stated below, Exceptions to the above-stated design basis for exterior building, openings in seismic Category I structures are: (1) the opening for the truck bay in the radwaste loading area of the Mechanlcal..

Electrical Auxiliaries Building (MEAB) and (2) the opening for the rail oar 4aoess in the sent f&el.cask loading area of the FHB. These areas are not protected from flooding because they do not have an4 safety-related systems and components located near or below the maximum flood level which is required to perform ahy essential function.

In addition, the two -areas are separated from the remalnd& of the building by wails which do not contain openings below the maximum water surface-elevation corresponding to their location.

The Tendon Gallery Access Shaftoover (TGAS) is-pro~~~ ... .. -. n-The safety-related equipment in the ECWIS is protected from the effects of the design basis flood, The porsonnel access doors on the west wall are piovided with watertight doors; all other doors and openings are above the flood level, The dividing walls and doors between the ECWIS compartmenits minimize the potential for the propagation of flooding from one compartment to another.The three malnt6nance Imookout panels in the. exterior walls of the Diesel-Generator Building .(DOB), which, are located below the maximum water surface elevqtion of 45,0 fL MSL, are watertight and designed for the hydrostatic forces. Bach kmockout panpl allows aoeoss to only one of the three separate compartments within the structure, hnd. only one panel may be removed at one time, The dividing walls between the compartments preclude propagation of flooding from oae compartment to (.. j, 'another.3.4-2 Revision 13 STPEGS UFSAR The maintenanoe knockout panels in the exterior wall.of the room, housing the componont cooling water heqt exchangers in the MEAB are located below the maximum steady-state water level shown I on Figure 3,4-2. These panels are watertight.

Since mechanistic effects fronfi the MCR breach need not be evaluated, there is adequate time to. replace the knockout panels for theý remaining flood events of conceni.All exterior seism'ic Category I building wall and slab surfaces below grade are waterproofed.

This conservatively protects the substructure of seismio Category I buildings from. .groundwater, which is expected.

to stabilize between El. 17 ft and 26 ft (1 to 10 ft below grade) after decommissioiaing of the dewatering system. No waterproofing Is provided on exterior wall or slab surfaces above grade to protect against the effeots, of surge-waver mn-up because of its short duration.

All construotionjoluts in exterior walls and slabs (except for localized areas of blockouts) are provided with waterstops to the maximum flood level for that location -and can withstand hydrostatic and hydrodynamlo effects, All seismic joints betwieen Category I structures contain dual 9-in, water stops bapable .of withstanding potential seismic: and hydrostatic effects. Cracks ia concrete are: mnimized by imposing strict QA and QC prooedures oh the quality of concrete and onstruction techniques.

• 'DfaiffY~-f6 pf6vid~d Wl cka dhliainhWF~kteihiafl6ddi~fo~rThfii~teu -ri1tt~h rn --flooding through the inadvertent introduction of water thrdugh these drains into sesmio Category I structures, The duct býnks are'sealed so as to prevent backflow into safety-related areas, The cable in the duct banks Is designed/specified for submerged installations..

Leakage from groumdwater Into the FHB is prevented by the use of on exterior wall and slab surfaces located below grade. Should groundwater inleakage occur, if fs handled by the ptunps in the FHB stnnp; the three-train compartment siunps, and the transfer cart area sump. For Unit 1 only, accumulated groundwater inieakage to the 64 degree tendon buttress area drains through a penetration in the RCB, tendon gallery outer will and is. collected Itn the tendon gallery sump.-UV o -and slab surfaoes located below grade, Should grgundwater leakage occur, iJ'will be collected in sumps, Discharge from non-radioactive sump,% are routed to.the reservoir via a circulating water discharge line, Potentially radioactive discharge is pumped to the Liquid Waste Processing System (LWPS).3.4,2 Analysts Procedures 3.4.2.1 Phenerena Considered in Deskn Load Calculations, For external flooding, the design basis events considered in design load calculations afe as described in Section 3.4.1.3.4,2'.2 Flood-Force Apxlicatlon.

The design flood conditions and elevations have been... determinedtfrom an analysis of the.phenomena discussed in Section 3.4,1,., 3.4-3 Revision 13 STPEGS PSFEAR In order to establish the controlling load conditions resulting from the embankrment breach, both Instantaneous surge wave runup as well as the longer term, quasi-steady-state -conditions were analyzed.

Tlhe wav6 runup condition conservatively assumes that the maximum total force perpendicular to the south face of the plant structures Includes a dynamic component in addition to the associated hydrostatic fproes, The quasi-steady state condition assumes that only the hydrostatic component cottributes to the.development of the total force for this case, The latter condition resulted im higher water surface elevatiotis and greater hydraulic loads on power block structures.

The vertioal buoyant loading condition is the force equal to the weight of water displaced by a structure.

The discussion of lateral and vertical loadings is presented in the following subsections.

Table 3.4-1 shows a summary of different lateral loadings at various locations aroundplant and ECP structures, caused by their respective controlling flood conditions, Procedures used to determine flood loadings are.Identified in Sections 3.4.2,2.1 and 3A42,2,2.3.4,2.2.1 Lateral Loading: 3.4.22.1,1 , Lateral LoadIh on the Power Block

-Thb analysis of the lateral force on the jower block structures considered both the instantaneous wave runup and the quasi-steady state conditions.

This analysis determined that the maximum total lateral force on the power.-

when the-maximum-water-:level isreaohed'durlng-the-quasi-steady-s'taer, condition; Table 3.4-1 shows the controlling lateral forces (hydrostatic).

exerted on different power block structures.

These lateral forces are treated as triangular loadings on a vertical surface, valring.at a rate of 62.4 lb/t2/ft of structure dep1h, The procedures used to determine the dynamic taid hydrostatic loadings for the above analysis condlitions are discussed below: 1, Dynamic Force The dynamio force on the south side of the power block structures is determined by application of.linear nioomentum principl.s, The flow from the MCOR is assumed to be normal to the south side of the power block structures.

Therefore, the dynamic force exerted on the structures can be exprdssed by the. following momentum equation (Ref, 3.4-2), where: V =dynamic force normal to plant structure P c density of flow_ flo-w-rate V 0 = velocity of flow The maximum value .ofpQv 0 during surge formation is calculated.

This is the contilbution of momentum flux to the dynamic foroo. The. contribution of the unsteadiness -of momentu-n field is insignificant.v

2. H-tydirostatic Force 3,4-4 Revision 1.3

( STPEGS UFSAR The lateral hydrostatic force is determined by the following equation (Ref. 3.4-2): Fyd = yj, 2 where: Friyd =hydrostatic force, lb/ft of width h , water depth, ft 7w unit weight of water; lb/&.i .3.4.2.2.1.2 Lateral on the ECGWIS and the South BCP Embament-The determination of the maximum lateral force on the ECWIS considered both instantaheous arid quasi-steady-state conditions.

The maximum total force on the ECWIS is a result tf the MCR embankment breaoh discussý6 in Section 2,4,4,2,2.

This force is the. result of a water elevation of 41.0-ft mean sea level during'the quasi-steady state condition, since te south BC? embanlcment crest elevation Is 34.0 A MSL, it would be overtopped by the flood wave resulting from the MOR embankment breach. The- south BCembankment Is designed to withstand the lateral free based on the maximum water'elevation resulting from.MCR embankment 1/4" breach, 3.4.2,2,2 YePgical.Loai.at The'roofs of seismic Category I struotureý are designed to withstand the ýweight of the accumulated PM1, assuming completely clogged drains (Section 2.4.2,3), Table 3.4-1 shows the elevations df maximum watersurfas e used for, buoyancy calculations, The maximum buoyant force is calculated by assuming that the granular backfill around the structures is completely saturated so that the buoyant force will- occur as. soon as water arives at the plant area.3.4.3.1 Proteotlon Foa=. Safetyorelated systems, components and structures are proteated such that the plant can achieve and maintain a safe shutdown condition and prevent unacceptable radiological releases to thle environment, In generealdhe plant layout arrangement on maxfiiz-ing thlepli-aldsl sra ion-f cf-doundant Fl-or diverse safety-related components and systems from each other and from nonsafety..related Items, Therefore, there is minimal effect on other systems or components which are required for safe shutdown of the plant or to mitigate the 6onsequence ofinternal flooding.

I., Where separation is not fhasible, other protection features are employed.

These protectioAi features (4.! include the following:

3.4-5 R~vision 13 CS1 CALC. NO. STPNOC0i3-CALC-006 E N E R C 0 N CALCULATION COVER SHEET REV.i PAGE NO. I of 42 Title: Dose Rate Evaluation of Reactor Vessel Water Levels Client: STP during Refueling for EAL Thresholds Project: STPNOC013 Item Cover Sheet Items Yes No I Does this calculation contain any open assumptions that require confirmation? (If 0 ED YES, Identify the assumptions)

_2 Does this calculation serve as an "Alternate Calculation"? (If YES, Identify the " Z design verified calculation.)

Design Verified Calculation No.3 Does this calculation Supersede an existing Calculation? (If YES, identify the superseded calculation.)

Superseded Calculation No.Scope of Revision: Added reference for reactor vessel head thickness, and updated calculations with new value (7.19 in).Removed any detector specific calculations so results can be applied to any detector at these locations.

Made several editorial changes.Revision Impact on Results: The dose rates for the cases with reactor vessel head attached are higher due to the reduction in head thickness.

Study Calculation-l Final Calculation

[Safety-Related

[E Non-Safety Related E](Print Name and Sign)Originator:

Andrew Blackwell

.7/ Date: 3/2jl /~l Design Verifier:

Curt Lindner Date:: 2 4 .Approver:

Marvin Morris Marvin Mor "i Date:

ENERCON xc E!rc '. l .p Evy dy CALCULATION REVISION STATUS SHEET CALC. NO. STPNOC13-CALC-006 REV... 1 PAGE NO. 2 of 42 CALCULATION REVISION STATUS REVISION DATE DESCRIPTION 0 02/07/2014 1 03/21/2014 Updated containment dimensions including reactor vessel head thickness.

Added more detail to calculations section. Made editorial changes.PAGE REVISION STATUS PAGE NO. REVISION PAGE NO. REVISION All 0 5-16,18,19,21-1 26,28,30,31,34-42 APPENDIX REVISION STATUS APPENDIX NO. PAGE NO. REVISION APPENDIX PAGE NO. REVISION NO. NO. NO.A All 0 B All 1 CALC. NO. STPNOC13-.... ENE;C CALCULATION CL-0 I E N E R C 0 N DEINCALC-006 xcellence

--Evy projec. Every DESIGN VERIFICATION PLAN AND

SUMMARY

SHEET REV. 1 PAGE NO. 3 of-42 Calculation Design Verification Plan: The calculation will be reviewed for correctness of inputs, design criteria, analyzed methods, and acceptance criteria.The stated objectives and conclusions will be confirmed to be reasonable and valid.Assumptions will be reviewed and confirmed to be appropriate and verified to be valid based on sound engineering principles and practices.(Print Name and Sign for Approval -mark "N/A " if not required)Approver:

Marvin Morris Marvin Morris 'ZZZ.,, Date: Calculation Design Verification Summary: The calculation has been designated as Safety Related as noted in the cover sheet.The calculation has been verified to be correct and performed using appropriate design inputs, assumptions, analytical methods, design criteria, and acceptance criteria.The conclusions developed in the calculation are reasonable, valid, and consistent with the purpose and scope.The assumptions are appropriate and valid.Based On The Above Summary, The Calculation Is Determined To Be Acceptable.

A (PrintName and Sign)Design Verifier:

Curt Lindner Date: Others: Date:

CALCULATION CALC. NO. STPNOC13-CALC-006 0 E-N E R CO N DESIGN VERIFICATION REV. 1 Fxcefience-Every project Every do.CHECKLIST PAGE NO. 4 of 42 Item CHECKLIST ITEMS Yes No N/A Design Inputs -Were the design inputs correctly selected, referenced I (latest .revision), consistent with the design basis, and incorporated in the X calculation?

Assumptions

-Were the assumptions reasonable and adequately described, justified and/or verified, and documented?

Quality Assurance

-Were the appropriate QA classification and requirements assigned to the calculation?

X Codes, Standards, and Regulatory Requirements

-Were the applicable 4 codes, standards, and regulatory requirements, including issue and addenda, X properly identified and their requirements .satisfied?

Construction and Operating Experience

-Have applicable construction and operating experience been considered?

X Interfaces

-Have the design-interface requirements been satisfied, including interactions with other calculations?

X Methods -Was the calculation methodology appropriate and properly applied to satisfy the calculation objective?

X Design Outputs -Was the conclusion of the calculation clearly stated, did 8 it correspond directly with the objectives, and are the results reasonfble X compared to.the inputs?Radiation Exposure -Has the calculation properly considered radiation exposure to the public and plant personnel?

...Acceptance Criteria -Are the acceptance criteria incorporated in the 10 calculation sufficient to allow verification that the design requirements have X been satisfactorily accomplished?

Computer Softwvare

-Is a computer programn or software used, and if so, are the requirements of CSP 3.02 met?COMMENTS: In accordance with CSP 3.02, MCNP5 and SCALE6.0 have been verified for use on ENERCON computers..

Al (Pri?,t Name and Sigt)Design Verifier:

Curt Lindner f / Date: 12 e Others: ./ Date: C (

Calc. No. STPNOCI3-CALC-006 , E N E R C C N CALCULATION SHEET Rev. 1 Page No. Page 5 of 42 Table of Contents Section Pa e 1 .Purpose and Scope ................................................................................................................................

8 2. Sum m ary of Results and Conclusion

.................................................................................................

8 3 .R efe re n c e s .............................................................................................................................................

9 4. A ssum ptions ........................................................................................................................................

10 5. Design Inputs ......................................................................................................................................

10 5.1 Fuel Assem bly Param eters ..........................................................................................................

10 5.2 Containm ent D im ensions ............................................................................................................

11 5.3 Core Isotopic Inventory

..............................................................................................................

12 5.4 M aterial Com positions

...........................................................................................

....................

14 6. M ethodology

.......................................................................................................................................

16 7 .C a lc u latio n s .........................................................................................................................................

17 7.1 Source Term s ..............................................................................................................................

17 7.2 M CNP M odel Core Hom ogenization

.....................................................................................

20 7.3 M CNP M odel Geom etry .............................................................................................................

21 7.4 M CNP Source Definition

............................................................................................................

30 7.5 M CNP Tally Specification

..........................................................................................................

31 7.6 M CNP M aterial Cards ................................................................................................................

32 7 .7 R e su lts .........................................................................................................................................

3 3 7.7.1 Results w ithout Head .....................................................................................................

34 7.7.2 Results with Head ...............................................................................................................

36 Appendix A -ENERCON Reference EM A ILS ....................................................................................

38 Appendix B -Electronic File Listing ..................

.. ...........................................................

41 Caic. No. STPNOC13-CALC-006 F ENE RCON CALCULATION SHEET Rev. 1 Page No. Page 6 of 42 List of Figures Fgur Pa. e Figure 7-1 ORIGEN-S Input Deck for MCNP Source Term Calculation

..............................................

18 Figure 7-2 X-Z VISED Plot of Reactor Vessel and Concrete Reactor Pit (No Head) ...........................

21 Figure 7-3 X -Z VISED Plot of Containm ent ..........................................................................................

22 Figure 7-4 X-Y VISED Plot of the Containment Geometiy at Radiation Monitor Level .....................

23 Figure 7-5 MCNP Model Surface Cards ...............................................

26 Figure 7-6 M CNP M odel Cell Cards (No Head) ...................................................................................

27 Figure 7-7 X-Z VISED Plot of Reactor Vessel and Concrete Reactor Pit (With Head) ........................

28 Figure 7-8 MCNP Cell Cards (With Head) ..................................................

29 Figure 7-9 M CNP Source D efinition Cards ............................................................................................

30 F igure 7-10 M C N P T ally C ards ..................................................................................................................

3 1 Figure 7-11 ANSI/ANS-6.1.1-1977 Gamma Flux to Dose Conversion Factors ........................................

31 Figure 7-12 M C N P M aterial C ards .............................................................................................................

32 Figure 7-13 Dose Rate versus Water Height Plot for no Head Configuration

......................................

35 Figure 7-14 Dose Rate versus Water Height Plot for with Head Configuration

...................................

37 C C C Calc. No. STPNOC13-CALC-006 CALCULATION SHEET Rev.Page No. Page 7 of 42 List of Tables Table Table 5-1 Design Input Fuel Assembly Parameters for Westinghouse Vantage 5 Fuel .........................

10 Table 5-2 Design Input Contaimnent D im ensions ..................................................................................

II Table 5-3 Design Basis Core Shutdown Source Term ...........................................................................

13 Table 5-4 SCALE Standard Compositions used in MCNP Model .........................................................

15 Table 7-1 Binned Total Core Source Term ............................................................................................

19 Table 7-2 Summary of Surfaces Used for M CNP M odels .................................................................

.24 Table 7-3 Dose Rate Response as a Function of Water Level for no Head Configuration (mrem/h) ........ 34 Table 7-4 Dose Rate Response as a Function of Water Level for Head on Configuration (mrem/h) ........ 36 Calc. No. STPNOC13-CALC-006 E EN E RCON CALCULATION SHEET Rev. 1 1ý wei en -.-V 2,Y ;o'jr*::.

[vv,;X -Y ,q Page No. Page 8 of 42 1. Purpose and Scope The purpose of this calculation is to evaluate dose rates as a function of water height in the reactor vessel during refueling operations in order to set Emergency Action Level (EAL) thresholds for core uncovery.The dose rates are calculated at the locations of the containment m6nitors RE-8055 and RE-8099 so that dose rate measurements by these devices can be used to estimate water level in the core, upon failure of other water level detection systems. This evaluation will calculate the dose rate at full core uncovery, as well as maximum water levels with a detectable dose rate response.

Since the scope of this calculation concerns uncovering the reactor core, the effects of future fuel element storage in the nearby Fuel Storage Pit are not analyzed, since it's effects are negligible in comparison.

The containment building, components within the building, and the reactor vessel and contents are modeled simplistically because only order of magnitude results are needed. As such, the dose rate results should be considered as reasonably representative of the magnitude of the actual dose rate only.2. Summary of Results and Conclusion The dose rate results for the configuration without the reactor vessel head and with the reactor vessel head are provided in Section 7.7.1 and Section 7.7.2, respectively.

The dose rate with the core uncovered (i.e. water at the top of the active length) is 2.23E+04 mrem/h with the head in place and 9.30E+06 mrem/h with the head removed. Detailed results of the dose rate as a function of water height are provided in Figure 7-13 with the head removed and Figure 7-14 with the head attached..C C Calc. No. STPNOC13-CALC-006 E N E R C 0 N CALCULATION SHEET Re. 1 Page No. Page 9 of 42 3. References

1. "Standard Composition Library," ORNL/NUREG/CSD-2/V1/R6, Volume 3, Section M8, March 2000.2. Calculation NC-65 10. "Core Radionuclide Inventory for Chapter 15 Accident Analysis." 3. RSICC Code Package CCC-750, "SCALE 6.0: A Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluation", Jan. 2009.4. "ORIGEN-S:

SCALE System Module to Calculate Fuel Depletion, Actinide Transmutation, Fission Product Buildup and Decay, and Associated Radiation Source Terms", I.C. Gauld, O.W.Hermann, & R. M. Westfall.

Jan. 2009.5. STPOOI-CPC-001.

Computer Program Certification MCNP5 Version 1.4 and SCALE 6.0.6. ENERCON email firom Paul Sudnak, dated December 9, 2013. (Appendix A).7. Drawing 6C-18-N-5006, Rev. 9. "General Arrangement Reactor Containment Building Plan at El. 68' 0" Area G." 8. Drawing 6C-18-9-N-5007, Rev. 6. "General Arrangement Reactor Containment Building Section A-A Area G." 9. Drawing 6C-18-9-N-5008, Rev. 8. "General Arrangement Reactor Containment Building Section B-B Area G." 10. RSICC Code Package CCC-730, "MCNP/MCNPX Monte Carlo N-Particle Transport Code System 12 Including MCNP5 1.40 and MCNPX 2.5.0 and Data Libraries.," January 2006.11. ANSI/ANS 6.1.1-1977, Neutron and Gamma Flux-To-Dose Conversion Factors.12. ENERCON email from Paul Sudnak, dated February 3, 2014 (Appendix A).13. Drawing L5-0 I EM 101, Rev. 1. "Closure Head General Assembly." 14. Drawing 1142E24. "Model 4XLR Reactor 173 in. I.D. Vessel." 15. Drawing 2C26-9-S-1004, Rev. 4. "Steel Reactor Containment Building Cylindrical Shell Liner Sects. And Dets. Unit N' 1 & 2." 16. Drawing 1211E6. "4 Loop Rapid XL Reactor General Assembly."

Calc. No. STPNOC13-CALC-006 , E NE R :C:O N CALCULATION SHEET Rev. 1 Page No. Page 10 of 42 4. Assumptions The following assumptions are used in the core uncovery dose rate calculation:

I. The core is homogenized based on the typical Vantage 5 fuel assembly dimensions, taking into account the fuel rods and space between. Any small variations in fuel parameters will have a negligible effect on containment dose rates.2. Any non-fuel hardware is ignored since the primaly self-shielding occurs in the fuel itself, and there may be some unknown streaming effects tlhirough the non-fuel hardware.This homogenization takes into account the water level when calculating the isotopic weight fraction and homogenized density.3. The source term for this evaluation is based on the fission product inventory at the time of shutdown.

Because there is no cooling time, the fuel gamma source term will predominate and the N-gamma and hardware activation can be neglected.

4. The compositions of the containment structure and components are based on the values in the SCALE standard composition library [1].5. The RE-8055 and RE-8099 monitors are assumed to be 5 feet above the 68 foot level in order to take into account the mounting device.6. The containment outer concrete thickness is modeled as 3 feet thick. Because the backscattering off the containment walls is due to the steel liner, this dimension has a negligible impact on dose rates near the reactor vessel.5. Design Inputs 5.1 Fuel Assembly Parameters The following fuel assembly parameters are used in the core homogenization in the MCNP model. They are based on typical fuel assembly values for Westinghouse Vantage 5 fuel.Table 5-1 Design Input Fuel Assembly Parameters for Westinghouse Vantage 5 Fuel Parameters Value Unit Reference Westinghouse Assumption 1 Fuel Type Vantage +# Fuel Rods per Assy 264 Assumption 1 Assembly Array 17x17 Assumption 1 Enrichment 4 wt% Assumption1 Density (% of theoretical) 0.95 Assumption 1 Fuel Pellet OD 0.3225 [in] Assumption 1 Fuel Rod Pitch 0.496 [in] Assumption 1 C C C Calc. No. STPNOC13-CALC-006 SE N ERCON CALCULATION SfEET Rev. 1 Page No. Page 11 of 42 Parameters Value Unit Reference Fuel Rod OD 0.374 [in] Assumption 1 Clad Thickness 0.0225 [in] Assumption 1 Guide Tube CD 0.482 [in] Assumption 1 Guide Tube Thickness 0.020 [in] Assumption 1# Guide Tubes 24 Assumption 1 Instrument Tube OD 0.482 [in] Assumption 1 Instrument Tube Thickness 0.020 [in] Assumption 1# Instrument Tubes 1 Assumption 1 Active Length 14 [ft] Assumption 1 5.2 Containment Dimensions The following dimensions are based on drawings of the STP containment building and equipment.

Some parameters are estimated using scaling when the drawings do not detail the exact dimension.

These estimations are only applied to dimensions that have a negligible effect on the core uncovery dose rate analysis.Table 5-2 Design Input Containment Dimensions Dimension:

ft. in cm reference Reactor Pressure Vessel Elevation at top of active fuel 28 2 858.52 [6]Elevation at head level platform 38 6.5 1174.75 [8]Elevation at full water level in refueling cavity 66 6 2026.92 [8]Closure head thickness 0 7.19 18.2626 [13]Reactor pressure vessel inside diameter at shell 0 173 439.42 [14]Height of reactor vessel from bottom of fuel to head level 742.95 Calculated Steam Generator Elevation at bottom of SG 38 4 1168.4 [9]Elevation at top of SG 105 9.875 3225.4825

[9]Total SG height 2057.0825 Calculated SG outer diameter 500 [7] Scaled Calc. No. STPNOC13-CALC-006

,'2 E N E R C O N CALCULATION SHEET Rev.Page No. Page 12 of 42 Dimension:

ft. in cm reference Active Fuel Active fuel bottom elevation 12 1 368.3 [9]Active fuel height 14 0 426.72 [14]Concrete Wall Lower Height 38 6.5 1174.75 [9]Upper Height 85 0 2590.8 [9]Overall Height 1416.05 Calculated Thickness 2 0 106 [7] Scaled Width 874.776 [7] Scaled Length 2499.36 [7] Scaled Steam Generators Lower Modeled Height 85 0 2590.8 [9]Upper Modeled Height 105 9.875 3225.4825

[9]Overall Modeled Height 634.6825 Calculated Diameter 500 [7] Scaled Containment Upper modeled height 153 0 4663.44 [8]Lower modeled height 68 0 2072.64 [8]Net Height 2590.8 Calculated Inner Diameter 149 111/4 4570 [15]Liner Thickness 0 0.375 0.9525 [15]Dome Inner Radius 74 11/8 2285 [15]Concrete Thickness 3 0 91.44 Assumption 6 C C 5.3 Core Isotopic Inventory Core isotopic activities are provided in Table 11 of [2]. The isotope specific activities are given in terms of Ci/MWt, which is converted to curies based on the total core thermal power of 4,100 MWt [2]. These calculations are performed in EXCEL spreadsheet STP.xlsx.

A table of the input values is shown in Table 5-3, below.C Calc. No. STPNOCI3-CALC-006 SENERC N CALCULATION SHEET Rev. I Page No. Page 13 of 42 Table 5-3 Design Basis Core Shutdown Source Term'Isotope Ci/MWt Ci Isotope Ci/MWt Ci Kr83m 3.41E+03 1.40E+07 Rul06 1.34E+04 5.49E+07 Kr85m 7.07E+03 2.90E+07 Rhl05 3.05E+04 1.25E+08 Kr85 2.93E+02 1.20E+06 Zr95 4.39E+04 1.80E+08 Kr87 1.34E+04 5.49E+07 Zr97 4.39E+04 1.80E+08 Kr88 1.90E+04 7.79E+07 Nb95 4.32E+04 1.77E+08 Kr89 2.32E+04 9.51E+07 La140 4.63E+04 1.90E+08 Xe131m 2.68E+02 1.10E+06 La141 4.62E+04 1.89E+08 Xe133m 1.66E+03 6.81E+06 La142 4.15E+04 1.70E+08 Xe133 5.37E+04 2.20E+08 Pr143 3.90E+04 1.60E+08 Xe135m 1.02E+04 4.18E+07 Nd147 1.73E+04 7.09E+07 Xe135 1.34E+04 5.49E+07 Am241 2.75E+00 1.13E+04 Xe137 4.63E+04 1.90E+08 Cm242 1.05E+03 4.31E+06 Xe138 4.39E+04 1.80E+08 Cm244 6.17E+01 2.53E+05 1131 2.59E+04 1.06E+08 Ce141 4.39E+04 1.80E+08 1132 3.71E+04 1.52E+08 Ce143 4.15E+04 1.70E+08 1133 5.37E+04 2.20E+08 Ce144 3.41E+04 1.40E+08 1134 5.85E+04 2.40E+08 Np239 5.12E+05 2.10E+09 1135 4.88E+04 2.OOE+08 Pu238 8.71E+01 3.57E+05 Sb127 3.05E+03 1.25E+07 Pu239 1.96E+01 8.04E+04 Sb129 8.29E+03 3.40E+07 Pu240 2.48E+01 1.02E+05 Te127m 4.32E+02 1.77E+06 Pu241 4.17E+03 1.71E+07 Te127 3.05E+03 1.25E+07 Rb86 9.92E+01 4.07E+05 Te129m 1.22E+03 5.OOE+06 Cs134 5.37E+03 2.20E+07 Te129 8.05E+03 3.30E+07 Cs136 1.54E+03 6.31E+06 Te131m 3.66E+03 1.50E+07 Cs137 3.17E+03 1.30E+07 Te132 3.82E+04 1.57E+08 Y90 3.56E+03 1.46E+07 Ba137m 2.93E+03 1.20E+07 Y91 3.41E+04 1.40E+08 Ba139 4.98E+04 2.04E+08 Y92 3.41E+04 1.40E+08 Bal40 4.63E+04 1.90E+08 Y93 3.90E+04 1.60E+08 Mo99 4.83E+04 1.98E+08 Sr89 2.68E+04 1.1OE+08 Tc99m 4.07E+04 1.67E+08 Sr90 2.37E+03 9.72E+06 Rul03 3.90E+04 1.60E+08 Sr9l 3.17E+04 1.30E+08 Rul05 2.68E+04 1.10E+08 Sr92 3.41E+04 1.40E+08 Ci = Ci/MWt x 4,100 MWt Calc. No. STPNOC13-CALC-006 E NE RCON CALCULATION SHEET Rev.1 Page No. Page 14 of 42 5.4 Material Compositions The following compositions used in the MCNP model are taken from the SCALE standard composition library [1] and are shown in Table 5-4.C C C Calc. No. STPNOC13-CALC-006"SI ENERC0N CALCULATION SHEET Rev. 1 Page No. Page 15 of 42 Table 5-4 SCALE Standard Compositions used in MCNP Model Material Isotope Weight Fraction Reference Zry- 4 Zr 0.9823 [1](6.56 g/crn 3) Sn 0.0145 Cr 0.0010 Fe 0.0021 Hf 0.0001 U0 2 U-235 0.0353 [1](10.412 g/Cml13)2 U-238 0.8461 O 0.1186'Air C 0.0001 [1](1.21E-03 g/crn 3) N 0.7651 O 0.2348 Water H 0.1111 [11 (0.9982 g/cm 3) 0 0.8889 SS-304 Fe 0.6838 [1](7.94 g/Cln) Cr 0.1900 Ni 0.0950 Mn 0.0200 Si 0.0100 C 0.0008 P 0.0004 Concrete 0 0.5320 [1](2.30 g//cin') Si 0.3370 Ca 0.0440 Al 0.0340 Na 0.0290 Fe 0.0140 H 0.0100 Carbon Steel C 0.0100 [1](7.82 g/clll) Fe 0.9900 2 Based on 95% of theoretical density, Assumption

1.

Calc. No. STPNOC13-CALC-006 V.3 EN ER CO N CALCULATION SHEET Rev. 1 Page No. Page 16 of 42 6. Methodology The reactor source terms are computed with ORIGEN-S of the SCALE 6.0 code package [3, 4].The ORIGEN-S decay sequence is used to bin design input isotope specific activities into energy dependent photon bills. These energy specific photon emission bins are used as input for the energy distribution described by the MCNP source definitions.

The ORIGEN-S sequence in the SCALE6.0 program package is verified for use in safety related calculations

[5]. The program certification form is maintained in the project file.MCNP5, release 1.40 [10], Monte Carlo transport is used to determine the dose rates. The ENDF/B-VI neutron cross section library, ENDF60, and the ENDF/B-VI Release 8 Photo-atomic Data gamma cross section library, MCPLIB04 are utilized in the transport computations.

This software has been verified for use in safety related calculations

[5].The detailed engineering drawings are converted into MCNP surface and cell cards in the proper dimensions.

The radiation monitors of interest are modeled as point detectors to determine the expected dose rate for those detectors.

The dose rates are calculated as a function of water height for two reactor refueling conditions:

1. With Head -the reactor is modeled with an 7.19 inch carbon steel plate as indicated in Table 5-2, which is additional attenuation between source and detector.2. Without head -the reactor is modeled with nothing between the active fuel zone and containment.

For low water levels, variance reduction is accomplished with a geometric importance map that is imposed on the homogenized core. Without significant amounts of water present, this is enough to calculate statistically sound dose rate results. Once the water depth reaches a height where the variance of the results reaches an unacceptable level, a superimposed weight windows mesh is utilized to improve the variance reduction of the simple geometric scheme. The weight windows are iteratively generated using the MCNP weight windows generator card with a mesh over the existing geometry.

All final dose rates presented in this calculation include weight windows variance reduction.

C C C Calc. No. STPNOCI3-CALC-006 3 E N E R C 0 N CALCULATION SHEET Rev. 1 Page No. Page 17 of 42 7. Calculations

7.1 Source

Terms In order to convert the isotope specific activity into an energy spectrum, ORIGEN-S of the SCALE6.0 code package is used to initiate a decay and bin into 19 photon energy groups. The energy groups along with their associated activities are used in the MCNP source definition to model the anticipated radiation emission following shutdown.The ORIGEN-S input deck, STPEAL.inp, is provided below in Figure 7-1. This input has a simple decay case where the inputted isotopic composition in curies is decayed. The isotope is specified in the 73$$card using the special identifier described in Section F7.6.2 of the ORIGEN-S manual, and the activity in curies is specified in the 74** card. The time steps for the decay are given on the 60** card in years.Although multiple time steps are calculated, the source term with zero decay time is used in this calculation to model the core immediately after shutdown.

The output of the decay is given in terms of photons/s/Energy-Group., which is automatically normalized in the MCNP input.

Caic. No. STPNOC13-CALC-006 E N E R CO N CALCULATION SHEET Rev. 1 Page No. Page 18 of 42 C Figure 7-1 ORIGEN-S Input Deck for MCNP Source Term Calculation

=origens 0as all 71 e t PWR Source Term STP ELA Analysis 3$$ 21 1 1 a4 27 a16 4 a33 19 e t 355$ 0 t 545$ a8 0 all 2 e 565$ 0 6 a6 1 alO 0 a13 66 5 3 0 2 0 e 57** 0 a3 1-16 e 9555 0 t STPEAL Ci Source Terms 60' 0 0.1 0.2 0.3 0.4 0.5 61"* 5rl-8 1+6 1+4 6555'GRAM-ATOMS GRAMS CUR: 3Z 0 1 0 1 3Z 1 1 1 1 3Z 1 1 1 1 815$ 2 0 26 1 e 825$ f2 83** 1.10E+07 1.00E+07 8.0(2.50E+06 2.OOE+06 1.6'4.00E+05 3.00E+05 2.0(84** 2.00E+07 6.43E+06 3.0'1.OOE+05 1.70E+04 3.0'3.05E+00 1.77E+00 1.3 3.25E-01 2.25E-01 1.0<-Call Origen-S Sequence<-Logical Unit Assignments-Binary Photon Library (71)<-Case Title<-Library Integer Constants-Units 83** Card Ci (4)-Gamma Energy Groups (19)<-Not Used<-Special Calculation Options-Cutoff Value (Default)-(0,n) Composition Dependent (-Subcase Control Constants-Decay Only Subcase (0)-Number of Time Intervals (6)-Number of Nuclides (66)-Unit of Time in Years (5)(-Not Used<-Not Used<-Subcase Title<-Subcase Basis<-Time (years)(-Cutoff Values (-Decay Period Print Triggers 6Z 6Z 6z (-Gamma Source Constants<-Produces Gamma Source Spectrum 3.OOE+06 (-Gamma Energy Groups 6.OOE+05 e 4.OOE+05 (-Neutron Energy Groups 1.OOE+01 (Not Used)4.OOE-01 1.OOE-05 e IES 0 0 0 1 1 1 0E+06 6E+06 0E+05 0E+06 0E+03 0E+00 0E-01 WATTS-ALL 1 0 0 6.50E+06 5.1.33E+06 1.1.OOE+05 5.1.85E+06 1.5.50E+02 1.1.13E+00 1.5.00E-02 3.WATTS-GAMMA 3Z 3Z 3Z.OOE+06 00E+06 00E+04 40E+06 00E+02 OOE+00 00E-02 4. OOE+06 8.OOE+05 1. 00E+04 9. 00E+05 3. OOE+01 8. 00E-01 1.OOE-02 735$ 360831 360851 360850 360870 360880 360890 541311 541331 (-Nuclide Identifiers 541330 541351 541350 541370 541380 531310 531320 531330 531340 531350 511270 511290 521271 521270 521291 521290 521311 521320 561371 561390 561400 420990 430991 441030 441050 441060 451050 400950 450970 410950 571400 571410 571420 591430 601470 952410 962420 962440 581410 581430 581440 932390 942380 942390 942400 942410 370860 551340 551360 551370 390900 390910 390920 390930 380890 380900 380910 380920 74** 1.40E+07 2.90E+07 1.20E+06 5.49E+07 7.79E+07 9.51E+07 1.IOE+06 <-Nuclide Concentrations (Ci)6.81E+06 1 .52E+08 1 .25E+07 S.90E+08 1. 80E+08 7 .09E+07 2. 10E+09 6.31E+06 9. 72E+06 7555 3 3 3 3 2.20E+08 4.18E+07 2.20E+08 2.40E+08 5.00E+06 3.30E+07 1.98E+08 1.67E+08 1.80E+08 i.77E+08 1.13E+04 4.31E+06 3.57E+05 8.04E+04 1.30E+07 1.46E+07 1.30E+08 1.40E+08 5. 49E+07 2 .00E+08 1. 50E+07 1. 60E+08 1. 90E+08 2. 53E+05 1. 02E+05 1. 40E+08 1. 90E+08 1.2 5E+07 I.57E+08 1.10E+08 1.89E+08 1.80E+08 1.71E+07 1. 40E+08 1. 80E+08 3.40E+07 1 .20E+07 5.4 9E+07 1.7 OE+08 1.70E+08 4.07E+05 1.60E+08 1. 06E+08 1 .77E+06 2 .04E+08 S.25E+08 1. 60E+08 1. 40E+08 2 .20E+07 1. 10E+08 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 (-Library Kind 2-Actinide 3-Fission Product 33333333333222333222223333333333 3 3 t 565$ fO t End IC Calc. No. STPNOC13-CALC-006 E ENE R CO0 N CALCULATION SHEET Rev.Page No. Page 19 of 42 The results of this calculation are summarized below in Table 7-1. These MCNP input source definition.

values will be used in the Table 7-I Binned Total Core Source Term Energy Energy Boundaries Photons/sec (MeV)1 0.01-0.05 9.29E+19 2 0.05-0.1 2.93E+19 3 0.1-0.2 6.54E+19 4 0.2-0.3 4,28E+19 5 0.3-0.4 1.52E+19 6 0.4-0.6 3.58E+19 7 0.6-0.8 4.35E+19 8 0.8-1 2.66E+19 9 1-1.33 1.29E+19 10 1.33-1.66 1.65E+19 11 1.66-2 5.57E+18 12 2-2.5 5.53E+18 13 2.5-3 1.98E+18 14 3-4 7.81E+17 15 4-5 3.48E+16 16 5-6.5 3.95E+11 17 6.5-8 1.75E+08 18 8-10 3.71E+07 19 10-11 2.01E+06 totals 3.95E+20 Calc. No. STPNOC13-CALC-006E N E R C 0 N CALCULATION SHEET Rev. 1..C , y dR Page No. Page 20 of 42 7.2 MCNP Model Core Homogenization.

Because the source term is given for the entire core, the self-shielding from the assemblies is an important part of the dose rate response in regions above the core. Particles born in the lower section of the core are very unlikely to penetrate through the core itself, and make it to the radiation monitors.

For simplicity, the core is modeled as a 3 dimensional cylinder with a uniformly distributed spatial particle distribution.

The calculations for the homogenization are done in the worksheet Compositions of the EXCEL workbook STP.xlsx.

A density and isotopic composition is calculated with the water level above the top of the fuel. A summary of the calculations for the core composition and density is shown below. The inputs are based on the dimensions in Table 5-1 and the compositions in Table 5-4.Rod Volume = 7r(Pellet Radius)2 x Active Length = (3.14)(0.16125 in)2 (168 in) = 13.7 in 3 Rod Massuo, = p x V = (10.96 cc) (0.95)(13.72 in 3) (2.54-) 2341.5 Number of Fuel Rods Assembly Massuo 2 = Rod Mass x Assembly R = (2341.5 g)(264) = 618.2 kg T (OD 2 ID 2 [(0.374 in)2 (0.329 in)2]Clad Volume = ---x Active Length = (3.14) -(168 in)4.17 in 3 9 (n3 cm, 3 Rod Masszry 4 p X V (6.56 (4.17 (2.54- n) = 448.7 g Number of Fuel Rods Assembly Masszry_4 = Rod Mass x = (448.7 g)(264) = 118.5 kg Assembly Assembly H 2 0 Volume = [(Assembly Width)2 -7r(Rod Radius)2 x 264] x Active Length= [(8.404 in)2 -(3.14)(06187 in)2 (264)](168 in) = 6993 in 3 Assembly MassH 2 o = p X V ( 0.9982+/-) (6993 in 3) (2.54-7n)

= 114.4 kg Assembly Volume = Active Length x (Assembly Width)2 = (168 in)(8.404 in)2 = 11865.4 in 3 Total Mass 1000(618.2

+ 118.5 kg + 114.4) kg Density = V = 4.38 g/cc (7 C (

Calc. No. STPNOC13-CALC-006 F-: ENERCoN CALCULATION SHEET Rev. 1 Page No. Page 21 of 42 7.3 MCNP Model Geometry The following MCNP model geometry is based on the containment dimensions summarized in Table 5-2.The model only focuses on the primary systems and components that provide shielding or reflection from the core to the radiation monitors.

These components include the reactor vessel, concrete in reactor pit, containment walls (reflection), and steam generators (reflection).

VISED plots of the model geometry are provided in Figure 7-2, Figure 7-3, and Figure 7-4. The MCNP surface cards with the model dimensions (cm) are shown in Figure 7-5, and the cell cards are shown in Figure 7-6 for the cases with no reactor head. A VtSED plot of the model with the reactor head is shown in Figure 7-7. The surface and cell cards for the cases with the reactor head are shown in and Figure 7-8, respectively.

Areas that are not of interest are given an importance of zero (white areas) so MCNP will not track particles in locations that will not contribute to the detector response.

A summary of surfaces used in constructing this geometry is shown in Table 7-2, including a description of macrobody dimensions.

Figure 7-2 X-Z VISED Plot of Reactor Vessel and Concrete Reactor Pit (No Head)* & ". c , ." " ... ..... , * " ..I-E~-----------

Air 806.5cc Reactor Pit Vessel Homogenized Core Calc. No. STPNOC13-CALC-006 L ENI R C N CALCULATION SHEET Re,. 1 Page No. Page 22 of 42 C Figure 7-3 X-Z VISED Plot of Containment 3 6488.75 cm (212' 10" Level)4295.14 cm (153' Level)1856.74 cm (73' Level)1704.34 cm (68' Level.)806.45 cm (38' 6.5" Level)3 Steam Generators are not full height. Also, they are not on the same X-Z plane as the core shown above. They are included for visualization purposes.

Caic. No. STPNOC13-CALC-006" E N E R CO N CALCULATION SHEET Rev. 1 Page No. Page 23 of 42 Figure 7-4 X-Y VISED Plot of the Containment Geometry at Radiation Monitor Level Radiation Monitor Steam Generators Radiation Monitor Cailc. No. STPNOCI3-CALC-006 i EN ER CO N CALCULATION SHEET Rev. 1 Page No. Page 24 of 42 Table 7-2 Suniniary of Surfaces Used for MCNP Models Surface Surface Type Number Dimensions Description RCC Xo Yo Zo Vx V, V, R 1 0 0 0 0 0 426.72 209.71 Active Fuel Region 2 0 0 0 0 0 700.45 219.71 Reactor Pressure Vessel Inner Surface 3 0 0 0 0 0 700.45 244.71 Reactor Pressure Vessel Outer Surface 31 0 0 700.45 0 0 18.26 244.71 Reactor Pressure Vessel Head 41 0 0 512.81 0 0 167.64 274.71 Concrete Void for*Primary Loop 42 0 0 512.81 0 0 167.64 411.71 Concrete Void for Primary Loop 10 0 0 700.45 0 0 106 244.71 Concrete Wall Cutout 11 444.71 843 700.45 0 0 2050 250 Steam Generator 1 12 444.71 843 720.45 0 0 2010 .230 Steam Generator Inner 1 13 -444.71 843 700.45 0 0 2050 250 Steam Generator 2 14 -444.71 843 720.45 0 0 .2010 230 Steam Generator Inner 2 15 -444.71 -843 700.45 0 0 2050 250 Steam Generator 3 16 -444.71 -843 720.45 0 0 2010 230 Steam Generator Inner 3 17 444.71 -843 700.45 0 0 2050 250 Steam Generator 4 18 444.71 -843 720.45 0 0 2010 230 Steam Generator Inner 4 21 0 0 1694.34 0 0 2600.8 2285 Containment Inner Liner Surface 22 0 0 1694.34 0 0 2600.8 2285.95 Containment Inner Concrete Surface 23 0 0 1694.34 0 0 2600.8 2377.39 Containment Outer Concrete Surface RPP -x x -Y Y -Z Z 4 -498 498 -498 498 -498 700.45 Concrete Surrounding RPV 8 -1250 1250 -437 437 806.45 2116.45 Concrete Wall Fuel Pit Inner 9 -1356 1356 -543 543 700.45 2116.45 Concrete Wall Fuel Pit Outer SPH X 0 Y 0 Zo R Calc. No. STPNOCI3-CALC-006E N E R CO N CALCULATION SHEET Rev. 1 Page No. Page 25 of 42 Surface Surface Type Number Dimensions Description 5 0 0 0 219.71 Bottom of Reactor Pressure Vessel Inner 6 0 0 0 244.71 Bottom of Reactor Pressure Vessel Outer 24 0 0 4295.14 2285 Containment Dome Inner Liner Surface 25 0 0 4295.14 2285.95 Containment Dome Inner Concrete Surface 26 0 0 4295.14 2377.39 Containment Dome Outer Concrete Surface Pz z 7 0 Fuel Bottom 71 700.45 Top of RPV 20 Variable Water Level 27 4295.14 Spring Line 28 1704.34 68' Level 101-110 42.672 426.72 Geometric Importance Divisions in Active Zone Calc. No. STPNOC13-CALC-006

~ EN ER CON CALCULATION SHEET Rev. 1 Page No. Page 26 of 42 C Figure 7-5 MCNP Model Surface Cards4 c surfaces 1 rcc 0 0 0 0 0 426.72 209.71 2 rcc 0 0 0 0 0 700.45 219.71 3 rcc 0 0 0 0 0 700.45 244.71 31 rcc 0 0 700.45 0 0 18.26 244.71 4 rpp -498 498 -498 498 -498 700.45 41 rcc 0 0 512.81 0 0 167.64 274.71 42 rcc 0 0 512.81 0 0 167.64 411.71 5 sph 0 0 0 219.71 6 sph 0 0 0 244.71 7 pz 0 71 pz 700.45 8 rpp -1250 1250 -437 437 806.45 2116.45 9 rpp -1356 1356 -543 543 700.45 2116.45 10 rcc 0 0 700.45 0 0 106 244.71 11 rcc 444.71 843 700.45 0 0 2050 250 12 rcc 444.71 843 720.45 0 0 2010 230 13 rcc -444.71 843 700.45 0 0 2050 250 14 rcc -444.71 843 720.45 0 0 2010 230 15 rcc -444.71 -843 700.45 0 0 2050 250 16 rcc -444.71 -843 720.45 0 0 2010 230 17 rcc 444.71 -843 700.45 0 0 2050 250 18 rcc 444.71 -843 720.45 0 0 2010 230 20 pz 365.76 21 rcc 0 0 1694.34 0 0 2600.8 2285 22 rcc 0 0 1694.34 0 0 2600.8 2285.95 23 rcc 0 0 1694.34 0 0 2600.8 2377.39 24 sph 0 0 4295.14 2285 25 sph 0 0 4295.14 2285.95 26 sph 0 0 4295.14 2377.39 27 pz 4295.14 28 pz 1704.34 101 pz 42.672 102 pz 85.344 103 pz 128.016 104 pz 170.688 105 pz 213.36 106 pz 256.032 107 pz 298.704 108 pz 341.376 109 pz 384.048 110 pz 426.72$$$$$$$$$$$$$$$$$$$$$$$$$$$$Active Fuel Region.Reactor Pressure Vessel Inner Surface Reactor Pressure Vessel Outer Surface Reactor Vessel Head Concrete Surrounding RPV Concrete Void for Primary Loop Concrete Void for Primary Loop Bottom of Reactor Pressure Vessel Bottom of Reactor Pressure Vessel Bottom of Active Zone Top of RPV Concrete Walls Fuel Pit Inner Concrete Wall Fuel Pit Outer Concrete Wall Cutout Steam Generator 1 Inner Steam Generator 1 Steam Generator 2 Inner Steam Generator 2 Steam Generator 3 Inner Steam Generator 3 Steam Generator 4 Inner Steam Generator 4 Water Elevation Surface Containment Inner Liner Surface Containment Inner Concrete Surface Containment Outer Concrete Surface Containment Dome Inner Liner Surface Containment Dome Inner Concrete Surface Containment Dome Outer Concrete Surface C$ Spring Line$ 68' Level$ Geometric Ia$ Geometric Ir$ Geometric Ir$ Geometric Ir$ Geometric Ir$ Geometric Ir$ Geometric Ir$ Geometric Ii$ Geometric Ii$ Geometric Ii aportance aportance nportance aportance mportance aportance mportance mportance mportance mportance Division Division Division Division Division Division Division Division Division Division Fuel Fuel Fuel Fuel Fuel Fuel Fuel Fuel Fuel Fuel Zone Zone Zone Zone Zone Zone Zone Zone Zone Zone 4 The surface cards for the MCNP model without the reactor vessel head does not have surface 31. The other surfaces are identical.

I C Calc. No. STPNOC13-CALC-006 ENERCON CALCULATIONN SEET Page No. Pagre 27 of 42 Figure 7-6 MCNP Model Cell Cards (No Head)c cells 101 1 -4.57 -1 ý101 102 1 -4.57 -1 101 -102 103 1 -4.57 -1 102 -103 104 1 -4.57 -1 103 -104 105 1 -4.57 -1 104 -105 106 1 -4.57 -1 105 -106 107 1 -4.57 -1 106 -107 108 1 -4.57 -1 107 -108 109 1 -4.57 -1 108 -109 110 1 -4.57 -1 109 -110 2 2 -0.9982 1 -3 #4 -20 4 4 -7.94 2 -3 7 -71 5 4 -7.94 5 7 #7 6 2 -0.9982 7 61 2 -0.9982 -20 71 (-10:-8)71 3 -1.21E-03

-42 41 7 5 -2.3 6 3 -4 #71 8 5 -2.3 8 -9 10 9 4 -7.94 -11 12 28 10 0 -12 28 11 4 -7.94 -13 14 28 12 0 -14 28 13 4 -7.94 -15 16 28 14 0 -16 28 15 4 -7.94 -17 18 28 16 0 -18 28 20 4 -7.94 21 -22 21 5 -2.3 22 -23 22 4 -7.94 24 -25 27 23 5 -2.3 25 -26 27 24 5 -2.3 28 9 #21 #22 11 13 15 17 30 3 -1.21E-03

(-24:-21:-8:-10:-2) 11 13 15 17 20 #8 #24 #2 1 999 0 1 #2 #4 #5 #6 #7 #71 #8 #9 #10#11 #12 #13 #14 #15 #16 #20 #21#22 #23 #24 #30 #61 imp: p=l imp: p=2 imp: p=3 imp:p=4 imp:p=8 imp:p=16 imp: p=32 imp :p=64 imp: p=128 imp:p=256 imp:p=256 imp: p=256 imp:p=256 imp:p=256 imp:p=256 imp:p=256 imp: p=256 imp:p=256 imp:p=256 imp: p=0 imp:p=256 imp:p=0 imp:p=256 imp: p=0 imp:p=256 imp: p=0 imp:p=256 imp:p=256 imp:p=256 imp:p=256 imp:p=256 imp: p=2 56$$$$$$$$$$$$$$$$$$$$$$$$$$$$$$Active Fuel Region Active Fuel Region Active Fuel Region Active Fuel Region Active Fuel Region Active Fuel Region Active Fuel Region Active Fuel Region Active Fuel Region Active Fuel Region Water Region RPV Shell Bottom RPV Shell Water Above Fuel Water. Above Vessel Head Void for Primary Loop Concrete Surrounding RPV Concrete above RPV Steam Generator 1 Inner Steam Generator 1 Steam Generator 2 Inner Steam Generator 2 Steam Generator 3 Inner Steam Generator 3 Steam Generator 4 Inner Steam Generator 4 Containment Liner Containment Wall Containment Dome Liner Containment Dome Concrete$ 68 foot level$ Air in Containment

$ Problem Boundary imp: p=0 Calc. No. STPNOCI3-CALC-006 , E N E R C O N CALCULATION SHEET Rev. 1 Page No. Page 28 of 42 C Figure 7-7 X-Z VISED Plot of Reactor Vessel and Concrete Reactor Pit (With Head)Reactor Head C CaIc. No. STPNOC13-CALC-006

'E N R Co N CALCULATION SHEET Rev.Page No. Page 29 of 42 Figure 7-8 MCNP Cell Cards (With Head)c cells 101 1 -4.57 101 102 1 -4.57 -1 101 -102 103 1 -4.57 -1 102 -103 104 1 -4.57 -1 103 -104 105 1 -4.57 -1 104 -105 106 1 -4.57 -1 105 -106 107 1 -4.57 -1 106 -107 108 1 -4.57 -1 107 -108 109 1 -4.57 -1 108 -109 110 1 -4.57 -1 109 -110 2 2 -0.9982 1 -3 #4 -20 31 4 4 -7.94 2 -3 7 -71 5 4 -7.94 5 7 #7 6 2 -0.9982 7 62 6 -7.8212 -31 61 2 -0.9982 -20 71 (-10:-8) 31 71 3 -1.21E-03

-42 41 7 5 -2.3 6 3 -4 #71 8 5 -2.3 8 -9 10 9 4 -7.94 -11 12 28 10 0 -12 28 11 4 -7.94 -13 14 28 12 0 -14 28 13 4 -7.94 -15 16 28 14 0 -16 28 15 4 -7.94 -17 18 28 16 0 -18 28 20 4 -7.94 21 -22 21 5 -2.3 22 -23 22 4 -7.94 24 -25 27 23 5 -2.3 25 -26 27 24 5 -2.3 28 9 #21 #22 11 13 15 17 30 3 -1.21E-03

(-24:-21:-8:-10:-2) 11 13 15 17 20 31 #8 #24 #2 1 999 0 1 #2 #4 #5 #6 #7 #71 #8 #9 #10#11 #12 #13 #14 #15 #16 #20 #21#22 #23 #24 #30 #61 31 imp:p =l imp:p=2 imp: p=3 imp: p=4 imp p =8 imp p=16 imp p=32 imp p= 64 imp p= 128 imp: p=256 imp:p=256 imp :p=256 imp:p=256 imp:p=256 imp:p=256 imp:p=256 imp:p=256 imp:p=256 imp:p=256 imp :p=256 imp:p=0 imp: p=256 imp:p=0 imp: p=256 imp:p=0 imp: p=256 imp:p=0 imp: p=256 imp: p=256 imp: p=256 imp:p=256 imp:p=256 imp:p=256 Active Fuel Region Active Fuel Region Active Fuel Region Active Fuel Region Active Fuel Region Active Fuel Region Active Fuel Region Active Fuel Region Active Fuel Region Active Fuel Region Water Region RPV Shell Bottom RPV Shell Water Above Fuel Reactor Vessel Head Water Above Vessel Head Void for Primary Loop Concrete Surrounding RPV Concrete above RPV Steam Generator 1 Inner Steam Generator 1 Steam Generator 2 Inner Steam Generator 2 Steam Generator 3 Inner Steam Generator 3 Steam Generator 4 Inner Steam Generator 4 Containment Liner Containment Wall Containment Dome Liner Containment Dome Concrete$ 68 foot level$ Air inside Containment

$ External to Problem imp:p=0 Cale. No. STPNOC13-CALC-006 NER C 0 , CALCULATION SHEET Rev. 1 Page No. Page 30 of 42 7.4 MCNP Source Definition The core source term is assumed to be uniformly distributed throughout the volume, and has an energy spectra based on the core inventory

[2]. Only the gamma source term is taken into account for this evaluation.

Because the source term is generated immediately after shutdown, the fuel gamma source term will predominate.

Therefore the N-gamma and hardware activation source terms can be neglected (Assumption 3). The source is defined on the MCNP sdef card using distributions to define the particle location and energy. The radius of the core is defined with the rad parameter, which automatically creates a uniform distribution based on a cylindrical geometry.

The ext and ays parameters define the direction and distance of the cylinder axis. These parameters combined define the core where the particles can be born. The erg parameter defines the energy spectrum of source particles and is based on the results of the ORIGEN-S calculation discussed previously.

This distribution is a histogram of energies represented by activities.

These are automatically normalized by MCNP to create a probability distribution.

The total activity is preserved in the tally multiplier.

The MCNP source definition cards are shown below in Figure 7-9. The sb card is a source biasing card, which in this case biases the particle generation to the upper end of the core. This is a variance reduction technique to improve the statistical certainty in the results.Figure 7-9 MCNP Source Definition Cards sdef rad=dl ext=d2 axs=0 0 1 erg=d8 (-Source Definition Card-Radius = dl-Extent = d2-Axis ý +Z-Energy = d8 sil 209.71 (-Core Radius Distribution si2 h 0 42.672 85.344 128.016 170.688 213.36 256.032 298.704 (-Core Axial Distribution 341.376 384.048 426.72 sp2 0 1 1 1 1 1 1 1 1 1 1 (-Actual Uniform Distribution sb2 0 0.001 0.001 0.01 G.01 0.01 0.1 0.1 0.1 1 1 (-Biased to Top Distribution c Fuel Gamma Spectra si8 h 1.000e-002 5.000e-002 1.000e-001 2.000e-001 3.000e-001 4.000e-001

(-Source Energy Groups 6.000e-001 8.000e-001 1.000e+000 1.330e+000 1.660e+000 2.000e+000 2.500e+000 3.000e+000 4.000e+000 5.000e+000 6.500e+000 8.000e+000 1.000e+001 1.100e+001 sp8 0.00E+00 9.288E+19 2.926E+19 6.537E+19 4.277E+19 1.521E+19 3.578E+19

(-Source Emission on Energy Basis 4.352E+19 2.66E+19 1.289E+19 1.649E+19 5.572E+18 5.527E+18 1.984E+18 7.812E+17 3.48E+16 3.947E+lI 1.75E+08 37100000 2009000 C C C Calc. No. STPNOCI3-CALC-006 ENERCON CALCULATION SHEET Rev.I Page No.Page 31 of 42 7.5 MCNP Tally Specification The tallies used in this evaluation are point detectors placed at approximate locations of radiation monitors RE-8055 and RE-8099. Point detectors are chosen because they use quasi-deterministic dose calculations that will provide better results than surface or cell based tallies that require the particles to enter those regions. The inputs to this card are the coordinates of the dose points followed by an exclusion zone (reduce variance), as well as a multiplier card, which represents the total core activity in photons/sec.

The tally cards are shown in Figure 7-10.Figure 7-10 MCNP Tally Cards f5c RE-8055 and RE-8099 f5:p -1200 -400 1909.24 20 1200 400 1909.24 20<-Tally Comment Card<-Tally 5 (point detector)x y z exclusion-1200 -400 1909.24 20 1200 400 1909.24 20<- Tally Multiplier (Total Activity)fm5 3.947E+20 In addition, the flux is multiplied by ANSI/ANS flux-dose conversion factors [11]. This is specified in MCNP using the de/df cards. These are shown in Figure 7-11.Figure 7-11 ANSI/ANS-6.1.1-1977 Gamma Flux to Dose Conversion Factors c- ------------------------------------------------------------------

c ANSI/ANS-6.1.l-1977 c Gamma Flux to Dose Conversion Factors c (mrem/hr)/(photons/cm2-s) c- ------------------------------------------------------------------

deC .01 .03 .05 .07 .10 .15 .20 .25 .30 .35 .40.45 .50 .55 .60 .65 .70 .80 1. 1.4 1.8 2.2 2.6 2.8 3.25 3.75 4.25 4.75 5. 5.25 5.75 6.25 6.75 7.5 9. 11.dfO 3.96E-03 5.82E-04 2.90E-04 2.58E-04 2.83E-04 3.79E-04 5.01E-04 6.31E-04 7.59E-04 8.78E-04 9.85E-04 1.08E-03 1.17E-03 1.27E-03 1.36E-03 1.44E-03 1.52E-03 1.68E-03 1.98E-03 2.51E-03 2.99E-03 3.42E-03 3.82E-03 4.01E-03 4.41E-03 4.83E-03 5.23E-03 5.60E-03 5.80E-03 6.01E-03 6.37E-03 6.74E-03 7.11E-03 7.66E-03 8.77E-03 1.03E-02 (-Energy Bins for Fluz to Dose Conversion

<-Energy Dependent Flux Multipliers Cale. No. STPNOC13-CALC-006 SE NER CO0N CALCULATION SHEET Rev.1 Page No. Page 32 of 42 7.6 MCNP Material Cards The MCNP material cards are provided in Figure 7-12. These are based on the compositions described in Table 5-4.Figure 7-12 MCNP Material Cards5 C ml 92235 -0.0245 92238 -0.5891 8016 -0.2521 40000 -0.1118 50000 -0.0017 24000 -0.0001 26000 -0.0002 1001 -0.0211 6012 -0.0001 m2 1001 2 8016 1 m3 6012 -0.000126 7014 -0.76508 8016 -0.234793 m4 6000 -0.0008 14000 -0.01 15031 -0.00045 24000 -0.19 25055 -0.02 26000 -0.68375 28000 -0.095 m5 26000 -0.014 1001 -0.01 13027 -0.034.20000 -0.044 8016 -0.532 14000 -0.337 11023 -0.029 m6 6012 -0.01 26056 -0.99$ Water$ Air$ SS 304$ Reg-Concrete

$ Carbon Steel.Material 1 composition will change based on the water level relative to the core. This only applies to water heights below 14 feet.

Calc. No. STPNOCI3-CALC-006 E N E R C 0 N CALCULATION SHEET Rev.Pagye No. Page 33 of 42 7.7 Results File NaminE Scheme: The MCNP input files are named with the following convention:

P-height-condition where: P = Project (STP)Height = water height from bottom of core (ft)Condition

= h -with head n -no head Calc. No. STPNOC13-CALC-006 E N E R C O N CALCULATION SHEET Rev. 1 Page No. Page 34 of 42 7. 7.1 Results without Head The dose rate as a function of water level is provided in Table 7-3 and plotted in Figure 7-13, below.Because the MCNP model geometry is symmetric in the x and y planes, the two point detector locations should provide the same dose rate. To increase the statistical certainty in the final result, the two individual dose rate responses and uncertainties are combined using inverse variance averaging.

All of the water levels described in the following sections refer to the level at the top of the fuel (i.e. 0 foot water level is at the top of the fuel assemblies and -13 feet is flange level).Table 7-3 Dose Rate Response as a Function of Water Level for no Head Configuration (mrem/h)Water Level (ft) Dose Rate 1 fsd Dose Rate 2 fsd Dose Rate Avg Avg fsd 0 9.27E+06 0.0081 9.34E+06 0.0109 9.30E+06 0.0065 2 4.26E+05 0.0078 4.31E+05 0.0093 4.28E+05 0.0060 4 2.31E+04 0.0236 2.32E+04 0.0247 2.32E+04 0.0171 6 1.73E+03 3.10E-02 1.69E+03 2.44E-02 1.70E+03 0.0192 8 1.51E+02 0.0302 1.51E+02 0.0287 1.51E+02 0.0208 10 1.40E+01 0.036 1.36E+01 0.0323 1.38E+01 0.0240 (C C Calc. No. STPNOCI3-CALC-006

.E N E R C O N CALCULATION SHEET Rev. 1 Page No. Page 35 of 42 Figure 7-13 Dose Rate versus Water Height Plot for no Head Configuration 1.0GE-07-1-OOE-35 i E Li .COE+04... ...... ............ ....... ..... .................

.. ......... ................

........ ........ ..........

........ ..... : ...... ....... .. ..........

... .....1.03E-O1 1.00E+ýDG 0 1 2 3 4 5 6 7 Water Levelfrom Top of Fuel (ft;8 9 10 1i Calc. No. STPNOC13-CALC-006 E N E .R CO N CALCULATION SHEET Rev.1 Page No. Page 36 of 42 7. 7.2 Results with Head The dose rate results for the cases with the head in place are the same, except the minimum detectable dose rate is lower due to the lower ambient dose rate in the containment.

The dose rates are listed in Table ,7-4 and plotted in Figure 7-14.Table,7-4 Dose Rate Response as a Function of Water Level for Head on Configuration (nireno/h)

Water Level (ft) Dose Rate 1 fsd Dose Rate 2 fsd Dose Rate Avg Avg fsd 0 2.16E+04 0.094 2.56E+04 0.185 2.24E+04 0.0838 2 1.87E+03 0.083 1.83E+03 0.074 1.85E+03 0.0554 4 1.11E+02 0.061 1.08E+02 0.069 1.10E+02 0.0455 6 8.89E+00 0.085 7.48E+00 0.048 7.82E+00 0.0418 8 8.95E-01 0.125 8.12E-01 0.093 8.42E-01 0.0742 C C Calc. No. STPNOCI3-CALC-006 SE N E R C 0 N CALCULATION SHEET RevN. P Page No. Page 37 of 42 Figure 7-14 Dose Rate versus Water Height Plot for with Head Configuration 1.00E40 L 4 ......l.COEiM;~ ~ ~~~ ~~~ ,0 E 0 --.. .. .... .... .... ..... ... -...... ..............

1.00E. 2 0 2 5 6 7 8 9 Water Level from Top of Fuel (ftl Calc. No. STPNOC13-CALC-006 FA E N E R C. 0 N CALCULATION SHEET Rev. 1 Page No. Page 38 of 42 Appendix A -ENERCON Reference EMAILS (7 Calc. No. STPNOC13-CALC-006 E N E R C 0 N CALCULATION SHEET e.Page No. Page 39 of 42 Drew Blackwell From. PaulSudnak Sent- Monday, December 09, 2013 9:55 AM Toe Chad Cramer Joanne Morris Cc, Marvin Morrn Jeff Ge.omatky Michael. Falkner, Jay Maisler Caleb Trainor

Subject:

RE STP Refueling Cavity Lever Cak Sure, let me fifnd .the elevation drawing for the cavIty. The "wrater level during refueling is the same water level as the.pool durinkg fuel transfer..

Th1e fuel is 2S"-2 in The level is and mid-loip is 32'-3t RC3 radiation mronitors (3R-0-055 and RE-OSS,) read from < I mRfhr to 2.S m-ihr during refueliing iflthe upper-internal pakarne or head are being rernredd, levels can increasýe to over 1XDImR/hr for the upper internals:

Levels on the refeli~ng deck ESS'0") Et rrid-loop vdill onyitncreese 10 mR/hrivth the vweterlevel that low. When the head is being de-tension ed by w'orker cn the head level platform (- 35' El.1, dose rtes at thsa location can read 50 to over Z10 mE/hr. The generalwarea dose rates from core radlation is usually les. than 100 unless there are lots of fuel leaks or high CIS corrosion and products-Dose rate- at the monitors a.t flange level are usually less than 5 mR/hr.Sudnale Enercon Ser.'ices Inc lZ906Tsmrnpa OaDsl, ulevard CukE 131 Te.mnpi1_

Terrace, Florida 33637 Office: (313) X9603 Fax: 4813) 9.62-1991 Cell: J813) 383-0950 Chad CrameN.-Sent. Friday, Der.eme 116, 2013 2:98 P31M Tb: loanne .Morri Cc: Marvin 3ef .S4Jdrak; Michal Falkner.Subjec ST 5"P -, eftielb 4 g Cavity Level Calct" Sficha. Faldmedbas.s the STP SFP ale and sent A to me for rmnie. I g-ikew Aih he an Jf GY."-romatzky and fftt0 indicated that he "7-Dd have ab-ity nov-er fli_ ne- Tre.dk or so pto do the refeling cai-ity lev'e calae Paul Calc. No. STPNOC13-CALC-006 ENERCON CALCULATION SHEET Rev.Page No. Page. 40 of 42 Dreiv Blackwell Fm: *Sudnak Paul <pJsudnak@STPEGS.COM>

Sent Monday. February 03.2014 3:44 PM To: Cateb Trainrso Drew BlackuelI C- Doral, Michae; jlay Maisler.Subject RE: Fuel Assembly Dimen.so.

on Thanks Ccaleb, ts.sum .e.l a only is from t&h matenEals bet.,-en the deterta and the cote. rDiseard reflieon I don't think'the SGts -ar betwaeen tf-oe and. detector n:or is the nal. Concrete should be hith :den3tyi.Atrrospbere shul1d b~e .saratdst~eam at then aspa"psi cortanmient spray rnftiat.on) and less 5; pzi fcOnrainmrna desi n pr-.rSe)lj insed vnth air.at the. original botainnment volur~ume atST..Tn detectors,.are Ion chamber-.

Derot includa nsutions.

The reactor VeSiel F-ad is aound "a' thick and ca rbon c-:teel. Pwillge you the actual drai;ino thi:knes., but I think itE3from the UFs " " 'rgarir, a pe-e from Mike oriJa?.Pa ul From: Caleb Trainor [raiftmdacb-ainor rcmn.crm]Sent: Monday, February 03, 2014 2:21 P14 To: Socnksak Paul; Dre~w E~ack-well.

CC-: Dorna, Mkcha-l; Iky Maisler SubjectR:

FFuel Assembly Dimension Drew is working on CS1/CG 1 where the concern is direct shine from the core due to lowered water levels and no fuel damage assiurned I think you may be thinking 6f the fission product ba frier calcs that I'm working on--Caleb From Sudnak, Paul <pisudnakgiPSTPEGS.CM>.'.

Sent: Monday, Fe-brury 3,2014 3:05 PM To: Caleb Trairoru, Drew Blackwaell Cc: ocrnai,, fM-iel; Jay Maisler.Subjec- RE: Fuet Dim'rension I think Caleb is'right here. Once is knimn, the dete,--bs are going Io .o the. gases Pprmarily albov- the: 68t Elevat-on, all threst wifllbe ritar attanetred by the concr-te fo ors,.inner and Bie-shield wall, learm genierstora, and the presarizer:

To mrodel all of those ztructure-would require an extferu-ise geormwtry and a cor.s&d÷able' Our intenshir2 oi~r' OW ElI andd tennine.the rcpotse.

dh Steam Generators, the inner and outer bio-shield bhe Pressurier.

Vtxth ant assumned homotrnau-mixc ba~scr on 2D- fuelrd'ar a, the dose rater should be argnmfcan:

Factoring in additional structures and -bevationswdl no tienmLi1zny chwillthr-ou t orme dGenclaa 0 mer-enc' wrlbedcladred.

Earlier today' I sent the lo.ation of the conTazmment hlgh range ron itos J73' E'cvatdon 5 trme thah'.ape the volurna of the teactor cnnta~nnzeirt building above the Ee Elcan Y-ou giVE. me a pser check h~erz7 Paul C C C Calc. No.STPNOCI 3-CALC-006"ENERCON CALCULATION SHEET Rev.1 Page No.Page 41 of 42 Appendix B -Electronic File Listing Volume in drive F is My Passport Volume Serial Number is IAEA-6007 Directory of F:\STPNOC013-CALC-006\Rev 1 03/14/2014 04:12 PM <DIR>03/14/2014 03/21/2014 02/06/2014 02/07/2014 03/14/2014 031/21/2014 03/14/2014 02/07/2014 03/14/2014 03/14/2014 03/14/2014 03/14/2014 04: 12 09:33 02 : 03 10:26 08: 44 09:32 04: 12 12:14 09:10 08:48 04:06 04:02 PM PM PM AM AM PM PM PM AM AM PM PM<DIR><DIR><DIR>0 100, 953 8,795 332,025 111,247 462,166 537,808 43,842 1,036,800 2,633,63 dir. dat EMAIL from Paul Sundak, Dec. 9 2013.pdf Inverse Variance Weighting.xlsx liner plate info.pdf mcnp origen RE Fuel Assembly Dimension.pdf RPV with core.pdf RPV.pdf STP.xlsx STPNOC013-CALC-006 Rl.doc 6 bytes 9 File (s)Directory of F:\STPNOC013-CALC-006\Rev 1\mcnp 03/21/2014 03/21/2014 03/21/2014 03/21/2014 02/06/2014 03/21/2014 03/21/2014 03/21/2014 03/21/2014 09:32 PM <DIR>09:32 PM <DIR>09:32 PM <DIR>09:32 PM <DIR>11:31 AM 09:34 AM 09:45 AM 09:45 AM 09:45 AM 5 File(s)head no head 137 STP.bat 18,720 STP.s: 4,053 STPdefault.sx 9,744 sx.log 2,007 sx.var 34,661 bytes Directory of F:\STPNOC013-CALC-006\Rev l\mcnp\head 03/21/2014 03/21/2014 03/12/2014 03/12/2014 03/12/2014 03/21/2014 03/21/2014 03/21/2014 03/21/2014 03/21/2014 03/14/2014 03/14/2014 03/14/2014 03/2-1/2014 03/21/2014 03/21/2014 03/21/2014 03/21/2014 03/13/2014 03/13/2014 03/13/2014 03/21/2014 03/21/2014 03/21/2014 03/21/2014 03/21/2014 03/12/2014 03/12/2014 03/12/2014 09:32 09:32 12: 51 08:50 08:50 09:45 04:41 09:17 04:41 09:17 08:27 03:42 03:42 09:45 04: 43 09:10 04:43 09:10 04:35 08:40 08:40 09:45 04:43 09: 17 04:43 09:17 01:17 08:51 08:51 PM PM PM PM PM AM PM PM PM PM AM PM PM AM PM PM PM PM PM PM PM AM PM PM PM PM PM PM PM<DIR><DIR>8,990 1, 104 924, 317 8,587 1,260 1,312 545,780 557, 996 8,990 1,260 942,029 8, 587 1,312 1,364 557,572 543,468 8,990 1,156 552, 616 8,587 1,260 1,312 551,487 565,735 8,989 1,104 966, 684 STP14h5 ST214h5m STPI4h5o STPI4h8 STP14hSm STPl4h8m2 STPl4h8o STP14h8o2 STP16h7 STPl6h7m STP16h7o STPI6h8 STP16h8m STPl6h8m2 STPI6h8o STP16h8o2 STP18h6 STPiSh~m STI18h6o STPI8h8 STP16h8m STPlSh8m2 STP18h8o STPIShSo2 ST P2Oh5 STP20h5m STP20h5o Calc. No. STPNOC13-CALC-006 F: E a j ERCON CALCULATION SHEET Rev. 1 Page No. Page 42 of 42 C 03/21/2014 03/21/2014 03/21/2014 03/21/2014 03/21/2014 03/12/2014 03/12/2014 03/12/2014 031/21/2014 03/21/2014 03/21/2014 03/21/2014 03/21/2014 09:45 AM 01:52 PM 07:40 PM 01:52 PM 07:40 PM 01:17 PM 08:51 PM 08:51 PM 09:45 AM 01:52 PM 07:07 PM 01:52 PM 07:07 PM 40 File(s)8,586 STP20h8 1,260 STP20h8m 1,364 STP20h8m2 550,735 STP20h8o 658,209 STP20h8o2 8,990 STP22h5 1,104 STP22h5m 936,997 STP22h5o 8,587 STP22h8 1,260 STP22h8m 1,364 STP22h8m2 547,911 STP22h8o 625,528 STP22h8o2 10,133,743 bytes Directory of F:\STPNOC013-CALC-006\Rev 1\mcnp\no head 03/21/2014 03/21/2014 03/13/2014 03/14/2014 03/14/2014 03/21/2014 03/21/2014 03/21/2014 03/13/2014 03/14/2014 03/14/2014 03/21/2014 03/21/2014 03/21/2014 03/13/2014 03/14/2014 03/14/2014 03/21/2014 03/21/2014 03/21/2014 03/13/2014 03/14/2014 03/14/2014 03/21/2014 03/21/2014 03/21/2014 03/13/2014 03/14/2014 03/14/2024 03/21/2014 03/21/2014 03/21/2014 03/13/2014 03/14/2014 03/14/2014 03/21/2014 03/21/2014 03/21/2014 09:32 PM <DIR>09:32 PM <DIR>04:57 PM 09:14 AM 09:14 AM 09:45 AM 03:46 PM 03:46 PM 04:57 PM 11:49 AM 11:49 AM 09:45 AM 03:46 PM 03:46 PM 04:35 PM 01:10 AM 01:10 AM 09:45 AM 02:02 PM 02:02 PM 04:56 PM 12:32 PM 12:32 PM 09:45 AM 03:46 PM 03:46 PM 04:56 PM 08:58 AM 08:58 AM 09:45 AM 03:46 PM 03:46 PM 04:56 PM 12:45 PM 12:45 PM 09:45 AM 03:46 PM 03:46 PM 36 File(s)8,270 STP14n7 1,364 STP14n7m 782,051 STP14n7o 8,468 STPI4n8 1,364 STP14n8m 447,366 STP14n8o 8,272 STP16n7 1,364 STP16n7m 702,157 STP16n7o 8,468 STP16n8 1,416 STP16n8m 437,587 STP16n8o 8,863 STP18n6 1,520 STP18n6m 541,013 STP18n6o 8,468 STPI8n8 1,572 STP18n8m 534,616 STP18n8o 8,268 STP20n7 1,364 STP20n7m 902,858 STP20n7o 8,467 STP20n8 1,312 STP20n8m 460,778 STP20n8o 8,273 STP22n7 1,364 STP22n7m 888,128 STP22n7o 8,468 STP22n8 1,312 STP22n8m 464,349 STP22n8o 8,272 STP24n7 1,364 STP24n7m 870,869 STP24n7o 8,468 STP24n8 1,312 STP24n8m 450,196 STP24n8o 7,599,621 bytes C Directory of F:\STPNOC013-CALC-006\Rev 1\origen 03/14/2014 03/14/2014 02/03/2014 02/04/2014 04:12 PM 04:12 PM 07:26 PM 02:43 PM<DIR><DIR>2,416 STPEAL.inp 99,996 STPEAL.out 102,412 bytes 2 File(s)

OPOP03-ZG-0009 Rev. 59 Page59 of 115 Mid-Loop Operation Addendum 1 RCS/RHR Simplified Elevation Diagram Page 1 of 1I REACTOR COOLANT SYSTEM PRESSURIZER SPILLOVER.34' 3.8" SECTION A-A HOT LEG STP D-0794 Rev 2 OPOP03-ZG-0009 Rev. 59 Page 60 of 115 Mid-Loop Operation Addendum 2 RVWL Sensor Elevations Page 1 of 1 NOTE" Top of Core is elevation 28 ft 2 inches." SG spillover is elevation 34 ft 3.8 inches.SENSOR UPPER HEAD PLENUM UNCOVERED INDICATED INDICATED SENSOR LEVEL DESCRIPTION LEVEL (%) LEVEL (%)All Covered 100 100 46' 4.75" Upper Head Full 1 64 100 45' 3.4" Upper Head Partially Drained 2 0 100 39' 4.9" Plenum Full 3 0 85 34' 10.1" Plenum NOT Full (Enter Reduced Inventory) 4 0 66 33' 5.5" Top of Hot Leg Nozzle 5 0 48 32' 3." Hot Leg Centerline 6 0 33 31'0.5" Bottom of Hot Leg Nozzle 70 20 30'1.6" Midway between Hot Leg Nozzle and Upper Core Plate 8 0 0 29' 2.7" Upper Core Plate STPEGS UFSAR TABLE 12.3.4-1 AREA RADIATION MONITORS Reactor Containment Building Tag Number and Location ( Range (mR/hr) (3) High Alarm Setpoint (mr/hr) (2)NI RA-RE-8052 10"1-10 4 1,000 Incore Instrumentation Room (-1 ft-6 in.)NIRA-RE-8053 10.1-104 100 Support across from elevator (-11 ft-3 in.)Ni RA-RE-8054 10"1-10 4 100 West Stair Landing (19 ft-0 in.)N1RA-RE-8055 101-104 100 North SG wall across from the head laydown area (68 ft-0 in.)NIRA-RE-8056 I0"-10 4 100 Support across from elevator (52 ft-0 in.)NIRA-RE-8099 10-1_10 4 100 South SG wall across from the in-containment fuel pool (68 ft-0 in.)1. Tag Number is shown for Unit 1. For Unit 2, change second position from I to 2.2. The alarm setpoints listed are typical and may be varied as necessary.

3. Range is based on microprocessor conversion factor and a detector signal which has a high degree of confidence.

Conversion factor will vary dependent on the detector calibration.

Exact ranges are found in plant instrument scaling manuals.12.3-24 Revision 15 STPEGS UFSAR.TABLE 12.3.4-1 (Continued)

AREA RADIATION MONITORS Fuel Handling Building Tag Number and Location () Range (nR/hr) High Alarm Setpoint (mr/hr) (2)NlRA-RE-8081 101-10 4 2.5-11 ft S of cols. 30.2 and S 5 (68 ft-0 in.)N1RA-RE-8084 101-10 4 2.5-24 ft S ofcols. 28 and T 5 (-21 ft-0 in.)NiRA-RE-8085 101-104 2.5-24 ft S of col. 28 and-6 ft E of col. S5 (-21 ft-0 in.)NIRA-RE-8086 10"1-101 2.5-24 ft S ofeol. 28 and-11 ft E of col. R 1 (-21 ft-0 in.)NlRA-RE-8087 10-1_104 2.5 col. 30.2 and 12 ft W of Col. R, (4 ft-0 in.)N1RA-RE-8088 10"-1_0 4 2.5* 3 ft S of col. 30.9 and col. R 1 (30 ft-0 in.)N1RA-RE-8089 i0 1-10 4 2.5 col. 28 and col. N (68. ft-0 in.)N1RA-RE-8090 i01-10 4 2.5 18 ft N of col. 30.2 and col. T 5 (68 ft-0 in.)NIRA-RE-8091 10-1i0 4 2.5 col. 34 and col. N (68 ft-0 in.)1. Tag Number is shown for Unit 1. For Unit 2, change second position from 1 to 2.2. The alarm setpoints listed are typical and may be varied as necessary.

12.3-25 Revision 15 STPEGS UFSAR TABLE 12.3.4-1 (Continued)

AREA RADIATION MONITORS Fuel Handling Building (Continued)

Range (mR/hr)10-2_107 Tag Number and Location (N1RA-RE-8097 33 ft S ofcols. 28 and 10 ft W of col. N (68 ft-0 in.)High Alarm Setpoint (mr/hr) (2)1,000 1. Tag Number is shown for Unit 1. For Unit 2, change second position from 1 to 2.2. The alarm setpoints listed are typical and may be varied as necessary 12.3-26 Revision 15 STPEGS UFSAR TABLE 12.3.4-1 (Continued)

AREA RADIATION MONITORS Mechanical Electrical Auxiliaries Buildin Tag Number and Location 0 Range (mR/hr) (3) High Alarm Setpoint (mr/hr) (2)N1RA-RE-8057 10-1&0* 0.5 col. 22 and -10 ft E of col. J (10 ft-0 in.)NIRA-RE-8058 10-1-10 2.5 col. 26 and col. J (10 ft-0 in.)NIRA-RE-8059 10"1-104 2.5 col. 27 and col G (10 ft-0 in.)N1RA-RE-8060 101-10 4 2.5-10 ft S of col. 30 and col. E (10 ft-0 in.)NIRA-RE-8061 10l-10 4 2.5-10 ft S of col. 24 and-11 ft W of col. E (10 ft-0 in.)NI RA-RE-8062 10-1-10 4 2.5-6 ft S of col. 31 and col. C (10 ft-0 in.)NI RA-RE-8063 10-I10 4 2.5-9 ft S of col. 28 and col. B (10 ft-0 in.)N1RA-RE-8064 I01-104 2.5-12 ft S ofcol. 24 and col. F (29 ft-0 in.)N1RA-RE-8065 10"1-104 2.5-5 ft N of col. 32 and col. C (29 ft-0 in.)1. Tag Number is shown for Unit 1. For Unit 2, change second position from 1 to 2.2. The alarm setpoints listed are typical and may be varied as necessary.

3. Range is based on microprocessor conversion factor and a detector signal which has a high degree of confidence.

Conversion factor will vary dependent on the detector calibration.

Exact ranges are found in plant instrument scaling manuals y, 12.3-27 Revision 15 STPEGS UFSAR TABLE 12.3.4-1 (Continued)

AREA RADIATION MONITORS Mechanical Electrical Auxiliaries Building (Continued)

Tag Number and Location 0 Range (mR/hr) ) High Alarm Setpoint (mr/hr) (2)N1RA-RE-8066 10-2_103 0.5-4 ft N of col. 22 and 14 ftE of col. C (35 ft-0 in.)NiRA-RE-8067 10-2_103 0.5 col. 22 and 10 ft E of col. J (35 ft-0 in.)NIRA-RE-8068 10i-10 4 2.5-10 RfN of col. 25 and col. 14 (41 ft-0 in.)N2RA-RE-8068 10"1-104 2.5-10 ft S of col. 24 and col. G (41 ft-0 in.)N1RA-RE-8069 10-2-10 3 0.5-12 ft S of col. 24 and-14 ft E of col. C (41 ft-0 in.)N1RA-RE-8070 10"2-103 2.5 col. 29 and col. C (41 ft-0 in.)NIRA-RE-8071 10"-10 3 2.5-18 ft S of col, 28 and 3 ft W of col. B (41 ft-0 in.)N1RA-RE-8072 10"1-104 .100-11 ftN ofcol. 29 and 5 ft W of col. D (41 ft-0 in.)N1RA-RE-8073 10"1-101 2.5 col. 29 and col. E (41 ft-0 in.)NIRA-RE-8074 10" 1_0 4 2.5-5 ft S of col. 31 and-7 ft W of col. C (41 ft-0 in.)1. Tag Number is shown for Unit 1. For Unit 2, change second position from 1 to 2.2. The alarm setpoints listed are typical and may be varied as necessary.

12.3-28 Revision 15 STPEGS UFSAR TABLE 12.3.4-1 (Continued)

AREA RADIATION MONITORS Mechanical Electrical Auxiliaries Building (Continued)

Tag Number and Location 1 Range (mR/hi') High Alarm Setpoint (mr/hr) (2)N1RA-RE-8075 10"-l101 15.0 col. 28 and -3 ft W of col. G (41 ft-0 in.)NIRA-RE-8076 102_103 0.5 col. 22 and -10 ft E of col. J (60 ft-0 in.)N1RA-RE-8077 10"1-104 2.5 col. 27 and col. J (60 ft-0 in.)N1RA-RE-8078 10"-10 4 15.0 col. 27 and col. F (60 ft-0 in.)N1RA-RE-8079 10-1_0 4 15.0 col. 25 and -2 ft W of col. F (60 ft-0 in.)N1RA-RE-8080 10-1_10 4 2.5 col. 26 and col. H (41 ft-0 in.)N1RA-RE-8082 10-1-104 2.5 col. 28 and -8 ft E of col. H (69 ft-0 in.)N1RA-RE-8083 10-1-104 15.0-10 ft S of col. 29 and 15 ft W of col. E (41 ft-0 in.)N1RA-RE-8098 102-107 1000-6 ft N of col. 25 and col. H (60 ft-0 in.)1. Tag Number is shown for Unit 1. For Unit 2, change second position from 1 to 2.2. The alarm setpoints listed are typical and may be varied as necessary.

12.3-29 Revision 15 Tag Number and Location )NIRA-RE-8092 col. 9 and col. P TGB (29 ft-0 in.)NiRA-RE-8093 col. 7 and col. M TGB (29 ft-0 in.)NIRA-RE-8094

-3 ft N of col. 23 and-14 ft W of col. B TSC-MEAB (72 ft-0 in.)STPEGS UFSAR TABLE 12.3.4-1 (Continued)

AREA RADIATION MONITORS Miscellaneous Buildings Range (mR/hr)10-2_103 10-2_103 10a-2107 High Alarm Setpoint (mr/hr) (2)0.5 0.5 1000 1. Tag Number is shown for Unit 1. For Unit 2, change second position from 1 to 2.2. The alarm setpoints listed are typical and may be varied as necessary.

12.3-30 Revision 15 STPEGS UFSAR TABLE 12.3.4-1 (Continued)

AREA RADIATION MONITORS Tag Number and Location (1)A1RA-RE-8050 RCB (68 ft-0 in.)C1RA-RE-8051 RCB (68 ft-0 in.)Post-Accident Monitors Range (R/hr)10°.108 10°_108 High Alarm Setpoint (R/hr) (2)2000 2000 1. Tag Number is shown for Unit 1. For Unit 2, change second position from 1 to 2.2. The alarm setpoints listed are typical and may be varied as necessary.

12.3-31 Revision 15 OPOP02-II-0002 Rev. 15 Page 13 of 17 RVWL Monitoring System Addendum 1 RVWL Sensor Elevations Page 1 of I NOTE* Top of Core is elevation 28 ft 2 inches.* SG spillover is elevation 34 ft. 3.8 inches.* IF the Delta T is between 25'F and 200'F, THEN RECORD the sensor is "wet" (covered with water) in the Wet/Dry colunim of the table on Data Sheet 1 or 2.IF the Delta T is GREATER THAN 200 0 F, THEN RECORD the sensor is "dry" (NOT covered with water) in the Wet/Dry column of the table on Data Sheet 1 or 2.Example: IF SENSOR No. 3 DRY is circled AND SENSOR No. 4 WET is circled, THEN the RVWL PLENUM INDICATED LEVEL (%) would be 85%, SENSOR Location 34' 10.1" and the LEVEL DESCRIPTION would be Plenum NOT Full.SENSOR No. UPPER HEAD PLENUM WET/DRY INDICATED INDICATED SENSOR LEVEL DESCRIPTION (circle one) LEVEL (%) LEVEL (%)All Wet 100 100 46' 4.75" Upper Head Full SENSOR No. 1 64 100 45' 3.4" Upper Head Partially Drained WET/DRY SENSOR No. 2 0 100 39' 4.9" Plenum Full WET/DRY SENSOR No. 3 0 85 34' 10.1" Plenum NOT Full WET/DRY SENSOR No. 4 0 66 33' 5.5" Top of Hot Leg Nozzle WET/DRY SENSOR No. 5 0 48 32' 3" Hot Leg Centerline WET/DRY SENSOR No. 6 0 33 31' 0.5" Bottom of Hot Leg Nozzle WET/DRY SENSOR No. 7 0 20 .30'1.6" Midway between Hot Leg Nozzle and WET/DRY Upper Core Plate SENSOR No. 8 0 0 29' 2.7" Upper Core Plate WET/DRY OPOP02-II-0002 Rev. 15 Page 15 of 17 RVWVL Monitoring System E Addendumn 3 RVWL Sensor Graphic Page 1 of I Elr T 34' Io.1"--3 -100 KLE 8LE33' 5.5" -4 32' 3" -'5 _. -50 31'0.5" -6 30'i .:6" --7 _29'2!.7"'..-8

-TOP OF _0 CORE T h NOZZLE h 28 21f... .t" "t " ". ... .." " CDI03516(06/26/03)

RVWL Sensor Graphic OPOP04-RC-0003 Excessive RCS Leakage Page 53 of 127 ndum9 ..Basis _ Basis Pag of 77_STEP DESCRIPTION FOR OPOP04-RC-0003 STEP 3.0 STEP: CHECK Trends For Any Of The Following Indications Of RCS Leakage:* Rad Monitor RT801 1 Particulate

-Rising* Reactor Coolant Drain Tank Level -Rising* Pressurizer Relief Tank Level -Rising* RCB Normal Sump Level -Rising PURPOSE: To determine if leakage is from RCS and not CVCS.BASIS: Indication of RT8011, RCDT PRT or RCB Normal Sump levels rising will confirm that the leakage is from RCS and not CVCS which is normally tied to the RCS.ACTIONS: Monitor trends from RT801 1, RCDT, PRT or RCB Normal Sump.INSTRUMENTATION:

Level indications located on CP004 and various plant computer monitors located in control room. Radiation Monitor Computer RM-1 1.CONTROL/EQUIPMENT:

N/A KNOWLEDGE:

N/A II- *II -I -4 This Procedure is ADDlicable in Modes 1, 2, 3, and 4

CG1 CALC. NO. STPNOC013-CALC-006 0 E'N E R C 0 N CALCULATION COVER SHEET REV.~he~te-Fvely P1610, L be dayj PAGE NO, l of 42 Title: Dose Rate Evaluation of Reactor Vessel Water Levels Client: STP during Refueling for EAL ThreshoIds Project: STPNOC013 item Cover Sheet Items Yes No 1 sDoes this calculation contain any open assumptions that require confirmation?. (If YES, Identify the assumptions)_____________________

2 Does this calculation serve as an "Alternate Calculation'? (if YES, Identify the design verified calculation.)

Design Verified Calculation No, 3 Does this calculation Supersede an existing Calculation? (If YES, Identify the superseded calculation.)

__ Superseded Calculation No.__ _ __ _ __ _ _Scope of Revision: Added reference for reactor vessel head thickness, and updated calculations with new value (7.19 In).Removed any detector specific calculations so results can be applied to any detector at these locations.

Made several editorial changes.Revision Impact on Results, The dose rates for the cases with reactor vessel head-Atacb-ed-ar.ehigher-d-utotbhe .reductionJn-head-

=tUilbk~~i-

-_.Study CalculationEr Final Calculation

[Safety-Related Z Non-Safety Related E (Print Name and Sign)Originator:

Andrew Blackwell Date: 3.211,'Design Verifier:

Curt Lindner Date: 14 Approver:

Marvin Morris Marvin Mord,,. .Date:

CAL.C.. NO,... STPNOCI3-CALC-006 N CALCULATION REVIS(ON STATUS SHET RE.GE NO. 2 CALCULATION MEISON S'TATUS.%REVISION DATE bH -RV I16.N 0 -02/0712014

,.0.1. 1 ....... ...... .ntiin .,,id ng reactor....... .,e .,,n.vesbe ieadbn thlknes. Added Mgre det a.l too ca* .'o .. .I ' taod hi .h ges.PAGE REV(ISION

ýSTATUS PAGE NO, REVISION.

PAGE NO.RVI All. 0.5-16,18,19,21-

.26,28,30,31,34-42

.!. ................

.i .... ........--

.... ...........

APPENDIX REVISION STATUS APPENDIX NO. PAGE NO REVISON APPENDI .PAGE NO. .REVISION NO. NO. NO.A All 0 B All.I C C.I CAMC. NO. STPNOC13-E N E R C O N CALCULATION CALC-006*p DESIGN VERIFICATION Excence--Ev, typ, oJe day PLAN AND

SUMMARY

SHEET REV. 1 PAGE NO. 3of42 Calculation Design Verification Plan: The calculation will be reviewed for correctness of Inputs, design criteria, analyzed methods, and acceptance criteria.The stated objectives and conclusions will be confirmed to be reasonable and valid.*Assumptions will be reviewed and confirmed to be appropriate and verified to be valid based on sound engineering principles and practices.(Print Name and Sign forApproVal

-mark "N/A" If not required)Approver:

Marvin Morris Marv nMo Date: Calculation Design Verification Summary: The calculation has been designated as Safety Related as noted in the cover sheet.The calculation has been verified to be correct and performed using appropriate design Inputs, assumptions, analytical methods, design criteria, and acceptance criteria.KT6n n T--01 6-- s- -e-d1 -1:hecl-s-are reasonable, valid, and consistent with. the purpose and scope.The assumptions are appropriate and valid.Based On The Above Summary, The Calculation Is Determined To Be Acceptable.(Print.bame and Sign)Design Verifier:

Curt Llndner /Date: ./ 1 Others. Date:

N C0CALCTATION ALNO. STPNOC13-CALC-006 ENSV:. C/ "O"N ..:..,.....D VERICATION R ,. .'"Y', I Y d/ C.'IST AGE NO. 4 of 42"Item "C ECKLIST ITEMS Yes No N/A S De~signiinpWts W e the design inputs o6irrectly selected.,.

1 (atest revisin), consistentWlith the design basis, and IIncorporated in the X .2 Assunptions

-Were. theasMuniptioxts reasonable ad adeqately described,.

2 .... e .,., *a .,, .,*ed ewifed, and do ete.? X. %Qufalityi Asu'ranc6 0ee ;the' appropriate QA olasificatioti and iegyirenents assigned to thd daiul ion? X' Codes, Staiidards, aid Regulatory Requirements-Were the applicable 4 codaoi, dand. regulhitoy' uiremeni; including ip.sue and.addefida, X prop~rlyldnif~iei snd~themi ,qt~Ar¢nient~saiisfled?.

.."" .:: ..: S Ojstr~ucftio and OpleritiiAguExperience 1- ve applicable b onsfructioi

.and operalihg experiiene beenh consil ered. .X* interface.

f :iave. the,. ini{nter ape nt&. be 'satisfied, 6 iclud n1 iftradtions other ealciii;ai;ons

.. X 7 M 4t8o-is Was the calculation' ioethodol g app r.priate an.. p:op r1 " applied to sisfy the calculato be" .Desig 'O'._uts-Was the conclusion of the oalculati6n stated, did 8 itco resoidndiretotly with the objecive'and "are the' X:

-1T~a" the-I almdafionxkpropqt ly i-/ierea idia n-i-e pc .......- -_ _ __ _ _A. ceptance Ciferia -. Are "tciteria

..in~orp6rted.

in the 10 caloulation:sufficient to allow verification That fth desigfi requiregerits have X been satisfactorily accomplished?.

ComPuter'Software

-Is a conlputerptrgii.

a o0 goflWare used, and if so, are the zequirements of CSP 3.02 imet?(1.C.,C COMMENTS: In acc6rdance with CSP 3.02, MCNP5 anmd SCALE6.0 have been verified for use on ENERCON comrnl

., .:. ..I Cale. No. STPNOC13-CALC-006 0 E NE R C0N CALCULATION SHEET Rev.Page No. Page 5 of 42 Table of Contents Section Pa~e 1. Purpose and Scope ...............................................................................................................................

8 2. Sum m ary of Results and Conclusion

...............................................................................................

8 3. References

.............................................................................................................................................

9 4. Assumptions

.......................................................................................................................................

10 5. Design Inputs ......................................................................................................................................

10 5.1 Fuel A ssembly Param eters ......................................................................................................

10 5.2 Containm ent D im ensions .........................................................................................................

11 5.3 Core Isotopic Inventory

..........................................................................................................

12 5.4 M aterial Compositions

.........................................................................................................

14 6. M ethodology

.....................................................................................................................................

16 7. Calculations

..........................................................................................................................................

17-7.1 Source Term s ..............................................................................................................................

17 7.2 M CNP M odel Core Hom ogenization

....................................................................................

20 7.3 M CNP M odel Geom etry ........................................................................................................

21 7.4 M CNP Source Definition

.....................................................................................................

30 7.5 M CNP Tally Specification

...................................................................................................

31 7.6 M CNP M aterial Cards ..........................................................................................................

32 7.7 Results .........................................................................................................................................

33 7.7.1 Results without H ead .....................................................................................................

34 7.7.2 Results with Head ..........................................................................................................

36 Appendix A -ENERCON Reference EM AILS .....................................

S...............................................

38 Appendix B -Electronic File Listing ..................................................................................................

.41 Calc. No. STPNOC13-CALC-006 C E NE R C 0 N CALCULATION SHEET Rev. 1 Page No. Page 6 of 42 List of Figures Figure Page Figure 7-1 ORIGEN-S Input Deck foK MCNP Source Term Calculation

.........................................

18 Figure 7-2 X-Z VISED Plot of Reactor Vessel and Concrete Reactor Pit (No.Head)

..............

21 Figure 7-3 X-Z VISED Plot of Containment

..............

,.......................................................

22 Figure 7-4 X:Y VISED Plot of the Contaiment Geometry at Radiation Monitor Level ......,...........

23 Figure 7-5 M CNP M odel Surface Cards..; ............................

2?.....6.........6........................

............................

26 Figure 7-6 M CNP M odel Cell Cards (No Head) ..........

....... ........ ...I..........

..........................

27 Figure 7-7 X-Z VISED Plot of Reactp0 Vessel and Concrete Reactor Pit (With Head),ý ................

28 Figure 7-8MCNP Cell Cards (W ith Head) ... ........ ...... ...... ..........

..........................

29 Figure 7-9 M CNP Sourcurc Defin fiion .....Card. ....... , ...... .........

.................

.............

30 Figure 7710 McNP Tally Cards .........................

........ .........

.............

..................

31 Figure 7-11 ANSI/ANS-6,1.1-1977 Gamma Flux to Dose Conversion Factors .. ...................................

31 Figure 7-12 M CNP M aterial Cards ... ... ...........

?...... ..........

..................

....... 32 Figure 7-13 Dose Rate versus Water 1eight Plot for no Head Configuration..;.....

..................

35 Figure 7-14 Dose Rate versus Water Height Plot for with 1ead Confguration................

.................

371/4'!C Cale. No. STPNOCI3-CALC-006 E N E R C 0 N CALCULATION SHEET Rev. 1 Page. No. Page 7 of 42 List of Tables Table Pane Table 5-1 Design Input Fuel Assembly Parameters for Westinghouse Vantage 5 Fuel ........................

10 Table 5-2 Design Input Containm ent Dimensions

.................................................................................

11 Table 5-3 Design Basis Core Shutdown Source Term ..........................................................................

13 Table 5-4 SCALE Standard Compositions used in.MCNP Model .........................................................

15 Table 7-1 Binned Total Core Source Term ...........................................................................................

19 Table 7-2 Summary of Surfaces Used for MCNP Models ..................................................................

24 Table 7-3 Dose Rate Response as a Function of Water Level for no Head Configuration (mrem/h) ........ 34 Table 7-4 Dose Rate Response as a Function of Water Level for Head on Configuration (mrem/h) ........ 36 Calc. No. STPNOC13-CALC-006 E K.CON CALCULATION SHEET Rev. 1-.Page No. Page 8 of 42 1. Purpose and Scope..The purpose of this calculation is to evaluate.dose rates as a function.

of watei hi6ihf in the reactor vessel during refu. eling d. , orat i6ns irn order. to set Eimrgenj y Ati~on. Level ;(E"t) tlhesholds foat core uncovery.The dose rates are calculated.

at the locations of the 'cointainmernt tnenitors REr8055 arid RE-8099 so that dose rate .neasurements by these devices can be usedto estiiiwate water level:in #e tore,iupon failure of other water level detection systems. Tis evaluatiwill calcula6teh:do:

rath at fu "eovery, as well as maximum water levels wiih -a detectable respIoIse.

Sincedthe scope 'of this calculation concerns the ffeqt of ture fuel elem.ent storage in thehearby."Fuel Storage Pit are.. not: analyzed, since it's effects. are negligible in comparison.

The dontainment building, components within the building, and the reactor vessel' and con/tents are modeledsimplistically because only order of magnitude

".esults ar.e needed. As such, the dogse rate results should be considered as reasonably reir'esentative of the magni tide of the actual dose rate only.2. Summary of Results suid. Conclusion The dose rate results for the .configuration without the: reactor vessel head and with the reactor vessel head are provided in Section 7.7.1 and Section 7.7.2, respectively.

The dose rate with the core uncovered (i.e. water. at the.top ofthie active length) is 2.23E+04 mrem/h with the head in place and 9.30E+06 mrem/h with the head ieimoved.

Detailed results. of the dose rate as a function of water height are provided in Figure 7-13 with the head removed and Figutrie 7-14 with the head attached.C C CIP Calc. No. STPNOC13-CALC-006ER 0N CALCULATION SHEET Rev. 1 Page No. Page 9 of 42 3. References

1. "Standard Composition Library," ORNL/NUREG/CSD-2/VI/R6, Volume 3, Section M8, March 2000.2. Calculation NC-65 10. "Core Radionuclide Inventory for Chapter 15 Accident Analysis." 3. RSICC Code Package CCC-750, "SCALE 6.0: A Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluation", Jan. 2009.4. "ORIGEN-S:

SCALE System Module to Calculate Fuel Depletion, Actinide Transmutation, Fission Product Buildup and Decay, and Associated Radiation Source Terms", I.C. Gauld, O.W.Hermann, & R. M. Westfall.

Jan. 2009.5. STP0O1-CPC-001.

Computer Program Certification MCNP5 Version 1.4 and SCALE 6.0.-6. ENERCON email from Paul Sudnak-,-dated Decemnber 9, 2013.. (Appendix A).7. Drawing 6C-18-N-5006, Rev. 9. "General Arrangement Reactor Containment Building Plan at El. 68' 0" Area G." 8. Drawing 6C-1 8-9-N-5007, Rev. 6. "General Arrangement Reactor Containment Building Section 7 A-A Area G." 9. Drawing 6C-18-9-N-5008, Rev. 8. "General Arrangement Reactor Containment Building Section B-B Area G." 10. RSICC Code Package CCC-730, "MCNP/MCNPX Monte Carlo N-Particle Transport Code System 12 Including MCNP5 1.40 and MCNPX 2.5.0 and Data Libraries," January 2006.11. ANSI/ANS 6.1.1-1977, Neutron and Gamma Flux-To-Dose Conversion Factors.12. ENERCON email from Paul Sudnak, dated February 3, 2014 (Appendix A).3_--0T T0 -" l r -l -'14. Drawing 1142E24. "Model 4XLR Reactor 173 in. I.D. Vessel." 15. Drawing 2C26-9-S-1004, Rev. 4. "Steel Reactor Containment Building Cylindrical Shell Liner Sects. And Dets. Unit No 1 & 2." 16. Drawing 121 1E6. "4 Loop Rapid XL Reactor General Assembly."

Catc. No. STPNOC13-CALC-006 G : NE -0 N CALCULATION SHEET Rev. 1 Page.No. Page 10 of 42 4. Assumptions The following assumptions are used in the core uncovery dose rate calculation:

1. The core is' homogenized based on the typical V.antage 5 fuel assembly dimensions, raking into accourif the ffel rods' nd spacebetween..

Ariy small Variations in fuel paramet6ers"will have a negligible qffect oni"containment doe raiesi.2, A rro~uo liadwar isignored since the prinmary self--shiel ingocrsi hefe*2. Any t n-onnfuel hhiadware.is g occurs in the fuel sitseland there.iai, besone uinkori aming effects through tle non-fuel hardware.This iakes'into'account the water level When.' 6 ui aing the isotopic weight fraction and honmogeniz6d density.3. The source term frthis is bdsedon thlefission proddut invento7r at the time-of shutdown.

BeeaiiS6 ffideivI.s -fibcooling time .the.fpqel gghmia .. o..u :rc6 term will predomiinate and the N-g'amma anid hard~afe activation~

can ben(~etd 4. The compositions of the containment structure and components are based on the values iii theSCALE st:aud:.cornpbsltioA lib .I [l]'5.. Thie:RE-8055 and RE-8099 assumed to be 5 feetabove the &8foot level in oidei to take into account th6d mounting device..6. The containment omiter concrete thickness is modeled as 3 feet thick. Because the ba"'.dkscattering off the containient wall§ is die to the steel liner,:"this dimension has a* negigible impat 6n t nearthe ýreatorvesseol.

5. Design Inpu:ts____

• _:"." 5.1 Fuel Assembly Parameters

.The following fuel assembly parameters are used in the core homogenization in the MCNP model. They are based on typical fuel assembly values for Westinghouse Vantage 5 fulel.C Table 5-1 Design Input Fuel Assembly Parameters for Westinghouse Vantage 5 Fuel Parameters Value Unit Reference Westinghouse Assumption I Fuel Type Vantage +#f Fuel Rods per Assy 264 Assumption 1 Assembly Array 17x17 Assumption 1 Enrichment 4 wt % Assumption 1 Density (% of theoretical) 0.95. Assumption 1 Fuel Pellet OD 0.3225 [in] Assumption 1 Fuel Rod Pitch 0.496 [in] Assumption I C:

Cale. No. STPNOC13-CALC-006 OEN ER CON CALCULATION SHEET Rev.E=4vue4v eypiojct Mvfydad________________

Page No. Page 11 of 42 Parameters Value Unit Reference Fuel Rod OD 0.374 [in] Assumption 1 Clad Thickness 0.0225 [in] Assumption 1 Guide Tube 01 0.482 [in] Assumption 1 Guide Tube Thickness 0.020 [in] Assumption 1# Guide Tubes 24 Assumption 1 Instrument Tube OD 0.482 [in] Assumption 1 Instrument Tube Thickness 0.020 [in] Assumption 1# Instrument Tubes 1 Assumption 1 Active Length 14 [ft] Assumption I 5.2 Containment Dimensions The following dimensions are based on drawings of the STP containment building and equipment.

Some parameters are estimated using scaling when the drawings do not detail the exact dimension.

These estimations are only applied to dimensions that have a negligible effect on the core uncovery dose rate analysis.Table 5-2 Design Input Containment Dimensions f=Dim-ension:...

..ft-- =in-- .......m ------reference-

--- -Reactor Pressure Vessel Elevation at top of active fuel 28 2 858.52 [6]Elevation at head level platform 38 6.5 1174.75 [8]Elevation at full water level in refueling cavity 66 6 2026.92 [8]Closure head thickness 0 7.19 18.2626 [13]Reactor pressure vessel Inside diameter at shell 0 173 439.42 [14]Height of reactor vessel from bottom of fuel to head level 742.95 Calculated Steam Generator Elevation at bottom of SG 38 4 1168.4 [9]Elevation at top of SG 105 9,875 3225A4825

[9]Total SG height 2057.0825 Calculated SG outer diameter 500 [7] Scaledý I -ýý'

Caic, No. STPNOC13-CALC-006 S N.ERCON CALCULATION SHEET Rev. 1 Page No. Page 12 of 42 C Dimension:

ft. in .cm reference Active Fuel :.Active fuel bottom elevation

.::12 1 368.3 .9]Active 14 0 426.72 [14]Low er H'eght e !ý k38s 6.5 1174.75 [9]upper eight, .85 0 2590.8:::

[9]Overeal Height .i416.05' .: Calcu'lated Thickness 2 0 .106,: [7] Scaled Width 8. 74.776: [7] Sealed Length 2499.36 [7] Scaled Steam Generators Lower Modeled Height 85. 0 2590.8 "91 Upper MogpeId Height,_.., .. ,. 105 9.875.: 3225.4825..9..

.....Overall.Modeled Height .. .,.634.6825, ,., Calcu1ated Diameter .7 ., ', 00 [7] Scaled Containment Upper modeled height: 153; 0 ý4663.44 [8]Lower modeled height 68 .,207264:

., [8]et-Hgit 1 d U t d25908 Clclat Inner Diameter .49... 11/4 --4570 , .*,1[']LlnerThlckness 0 0,,:0 375 .0.9525 [15]Dome Inner Radius .74 n1./s8 .2285 [151 Concrete Thickness 3 A 0 91.44 Assumption 6 AC-5 5.3: Core Isotopic Ivento .i .Core isotopic activities are provided in Table 11 of [2]. 'The isotope.specific activities are given in terms of Ci/MWt, which is converted to curies based on the total core thermal power of 4,100 MWt [2]. These calculations are performed in EXCEL spreadsheet STP.xlsx.

A table of the input values is shown in Table 5-3, below..C.

Calc. No. STPNOC13-CALC-006 E N E R C O N CALCULATION SHEET Rev. 1 Page No. Page 13 of 42 Table 5-3 Design Basis Core Shutdown Source Termi Isotope Ci/MWt Ci Isotope Ci/MWt CI Kr83m 3.41E+03 1.40E+07 Rul06 1.34E+04 5,49E+07 Kr85m 7.07E+03 2.90E+07 Rhl05 3.05E+04 1,25E+08 Kr85 2.93E+02 1.20E+06 Zr95 4.39E+04 1.80E+08 Kr87 1.34E+04 5.49E+07 Zr97 4,39E+04 1.80E+08 Kr88 1.90E+04 7.79E+07 Nb95 4.32E+04 1.77E+08 Kr89 2.32E+04 9.51E+07 La140 4.63E+04 1.90E+08 Xe131m 2.68E+02 1.10E+06 La141 4.62E+04 1.89E+08 Xe133m 1.66E+03 6.81E+06 La 142 4.15E+04 1.70E+08 Xe133 5.37E+04 2.20E+08 Pr143 3.90E+04 .1.60E+/-08 Xel35m 1.02E+04 4.18E+07 Nd147 1.73E+04 7.09E+07 Xe135 1.34E+04 5.49E+07 Am241 2.75E+00 1.13E+04 Xe137 4.63E+04 1.90E+08 Cm242 1.05E+03 4.31E+06 Xe138 4.39E+04 1.80E+08 Cm244 6.17E+01 2.53E+05 1131 2.59E+04 1.06E+08 Ce141 4.39E+04 1.80E+08 1132 3.71E+04 1.52E+08 Ce143 4.15E+04 1.70E+08 1133 5.37E+04 2.20E+08 Ce144 3,41E+04 1.40E+08 1134 5.85E+04 2.40E+08 Np239 5.12E+05 2.10E+09 1135 4.88E+04 2.OOE+08 Pu238 8.71E+01 3.57E+05 Sbb12-7= A3, 05 E03 E+04=Sb129 8.29E+03 3.40E+07 Pu240 2.48E+01 1.02E+05 Te127m 4.32E+02 1.77E+06 Pu241 4.17E+03 1.71E+07 Te127 3.05E+03 1.25E+07 Rb86 9.92E+01 4.07E+05 Te129m 1.22E+03 5.OOE+06 Cs134 5,37E+03 2.20E+07 Te129 8.05E+03 3.30E+07 Cs136 1.54E+03 6.31E+06 Tel31m 3.66E+03 1.50E+b7 Cs137 3.17E+03 1.30E+07..

Te132 3.82E+04 1.57E+08 Y90 3.56E+03 1.46E+07 Ba137m 2.93E+03 1.20E+07 Y91 3.41E+04 1.40E+08 Ba139 4.98E+04 2.04E+08 Y92 3.41E+04 1.40E+08 Bal40 4.63E+04 1.90E+08 Y93 3.90E+04 1.60E+08 Mo99 4.83E+04 1,98E+08 Sr89 2.68E+04 1.1OE+08 Tc99m 4.07E+04 1,67E+08 Sr90 2.37E+03 9.72E+06 Rul03 3,90E+04 1.60E+08 Sr9l 3.17E+04 1.30E+08 Ru105 2.68E+04 1.1OE+08 Sr92 3.41E+04 1.40E+08 Ci = Ci/MWt x 4,100 MWt Cale. No. STPNOC13-CALC-006 N " ERC ON CALCULATION SHEET ev. 1" ..-." ____. * .'. .. __.. _. , Page No. Page 14 of 42 5.4 Material Compositions.

The following compositions Used in the MCNP model are taken fromithe SCALEstandard composition library [1] and are shoWn in Table 5-4.C AM as..

Calc. No. STPNOC13-CALC-006 E N E R C: 0N CALCULATION SHEET Rev. 1 Page No. Page 15 of 42 Table 5-4 SCALE Standard Compositions used in MCNP Model Material Isotope Weight Fraction Reference Zry- 4 Zr 0.9823 [1](6.56 g/rn 3) Sn 0.0145 Cr 0.0010 Fe 0.0021 Hf 0.0001 U0 2 U-235 0.0353 [1](10.412 g/Cr 3)2 U-238 0.8461 O 0.1186 Air C 0.0001 [1](1.21E-03 g/cm 3) N 0.7651 o 0.2348 Water H 0.1111 [1](0.9982 g/cm 3) 0 0.8889 SS-304 Fe 0.6838 [1](7.94 g/cr 3) Cr 0.1900.I Ni 0.0950*I. ~ -. ---.-.------..

I. -- .-. I Mn 0.0200 Si 0.0100 C 0.0008 P 0.0004 Concrete 0 0.5320 [1](2.30 Wcm 3.) Si 0.3370 _Ca 0.0440 Al 0.0340 Na 0.0290 Fe 0.0140 H 0.0100 Carbon Steel C 0.0100 [11 (7.82 g/cm 3) Fe 0.9900 2 Based on 95% of theoretical density, Assumption

1.

JCai. No. STPNOC13-CALC-006 I .E.N E. E COýN CALCULATION SHEET Re.* ....-, ....Rev. 1._.__ ._..___...-._.

__._... _..._..._.__.._..

....__,.jPage N o. Page 16 of 42 0*Y. .. .* ...6. Methodl0 gy...The reactor source tehns are comp uted with ORJOEN-S of the SCALE 6.0 code package .3, 41.The 'ORIOEN-S decay' sequhce is used to bin design.input isotope specific t iavities..

into energy dependent photn bifis; These'.energy specific'.photon emission bins are use" as input fr the energy distribution described by the MCNP soutfce definitions.

The ORIGEN-S sequence in th e SCALE6.0 program package is. verified for use in safety related calculations

[5]. Thez program'.

ertification forhm ism'aintained in the project file.: MCNP5, release"l.40

[ 0], Monte' Carld transport is used to determine the. dose rates. The ENDF/B-VI neutron 'crss.s~etion librd.q, ENDF60 and th6" Release"8 Phtot-atomic:Data gamma cross section library,.MCLIB04 ar. utitlizd i the t cansit.iomputations.

Ths-sft-'are"hsbeen verified for use in safety relate:d 'calcuigtons

[51].." The detailed engineering drawings are convered ino MCN surface and cell cards in the proper dimensions..

The' radiation

'monitors of interest are modeled as point detedtors to .determind the expected dose rate f6r those detectors.'

The'dose.

rates are calculated as 'a functidn of water hdeigh for two reactor refueling conditions:

'1. With Head -the reactor is modeled With an 7.19 inch carbon'steel plate as indicated in Table 5-2,' whichis additional attenuation b6etween 'source and

2. Without head -the' reactor is modeled with :nothing between the active fuel zone and containment.

'For low water levels, variance redauction is!'accomplished with a geometric importance map that is imposed on the homogenized core. Without sig ificant armounts of water present,'

this is enough to calculate statistically souhd dose. rate resultsi; Once the water depth reaches a height where the variance of the results reaches ant unacceptable level, a superimposed weight windows mesh'is utilized to improve the variance reduction of the simple geometric scheme. The, weight wind0ows are iteratively generated using the MCNP weight windows generator card with a rmesh 'over the existing geometry.

All final dose rates presented in this calculation include Weight windows variance reduction.

.I%4 Calc. No. STPNOC13-CALC-006 OENER CON CALCULATION SHEET Rev. 1 Page No. Page 17 of 42 7. Calculations

7.1 Source

Terms In order to convert the isotope specific activity into an energy spectrum, ORIGEN-S of the SCALE6.0 code package is used to initiate a decay and bin into 19 photon energy groups. The energy groups along with their associated activities are used in the MCNP source definition to model the anticipated radiation emission following shutdown.The ORIGEN-S input deck, STPEAL.inp,.

is provided below in Figure 7-1. This input has a simple decay case where the inputted isotopic composition in curies is decayed. The isotope is specified ih the 73$$card using the special identifier described in SectionF7.6.2.

of the ORIGEN-S manual, and the. activity in curies is specified in the 74** card. The time steps for the decay are given on the 60** card in years.Although multiple time steps are calculated, the source term with zero decay time is used in this calculation to model the core immediately after shutdown.

The output of the decay is given in terms of photons/s/Energy-Group, which is automatically normalized in the MCNP input.

Cale. No. STPNOC13-CALC-006 D EN k C 0 N CALCULATION SHE ET Rev.Pa ge No. Page 18 of 42 I Figure 7-1 ORIGEN-S Input Deck for MCNP Source Term Calculation

=origens <-Call Origen-S Sequence 05$ all 71 e t <-Logical Unit Assignments-Binary Photon Library (71)PWR Source Term STP ELA Analysis <-Case Title 3$$ 21 1 1 a4 27 a16 4 a33 19 e t .,<-Library Integer Constants-ýUnitýs .83*.* Card Ci (4)* ~. ,-:Gamma Energy Groups (19)35$$ 0 t I.N" 'Used 545$ a8 0 al' 2 e '.Special aiculation Options-Cutoff Value (Default)-(an) Composi'tion Dependent 56$$ 0 6 a6 I alO 0 a13 66 5 3 0 2 0 e <-Subcase Control Constants"... -Decay Only Subcase (0)-Number of Tike Intervals (6).of.1Nclides (66)-Unit of Time in Years (5)57*4 0 3 1-16 e: '.... "...".." " " .ot Used.: 95$* 0 .(-Not used STPE34L (--SuboaSe, Title: Ci Source Terms A"SUbcase Bais 60** 0 0.1 0.2 .03 0.4 0.5 ..Time.(years) 61** 5ri-8 1+6 1+4 <-Cutoff Values 65$$ -.. Decay Pe iod Print Triggers'GRAM-ATOMS GRAMS CURIES WATTS-ALL WATTS-GAMMA 3Z 0 1 0 0 0 1 00 3Z 6z.3Z 1 11 1 .0 1 1 1 1 32 6z 3Z 1 1 1 1 1i. 11 32 6Z 815$ 2 0 26 1 e <-Gazima Source Constants 825$ f2. ... " Produces Gamma Source Spectrum 83** 1.10E+07 1.00E+07 8.002+06 6.50R+06 5.00E+06 4.002+06 3.00E+06 <-Gamma Energy Groups 2.50E+06 2.00E+06 1.66E+06 1.33E+06 1.DOE+06 8.00t+056.00E+05 4.OOE+05 3.00E+05 2.00E+05 1.00E+05 5.00E+04 1.00E+041e..

84** 2.00E+07 6.43E+063.O00E+06.1.852+06 1.40E+06 9.00E+05 4:.002+05

<-Neutron Energy Groups....4 002E305=1S=

--. 3 _3_5-5___021-OE÷0

___3.05E+00 1.77E+00 1.302+00 II3E+00 1.00E+00 8.00.E01 4.00E-01.3.25E-01 2.25E-01 1.00E-01 5.00F-02 3.00E-02 1.00E-02 1.OOE-05 e 735$ 360831 360851 360B50 360870 360880. 360890 541311 541331 <-Nuclide Identifiers 541330 541351 541350 541370 541380 531310 53'1320 531330 531340 531350 511270 511290 521271 521270 521291 521290 521311 521320 561371 561390 561400 420990 430991 441030 441050 441060451050 400950 450970 410950 571400 571410 571420 591430 601470 952410 962420 962440 581410 581430 581440 932390 942380 942390 942400 942410 370860 551340 551360 551370 390900 390910 390920 390930 380890 380900 380910. 380920 74** 1.40E+07 2.90E+07 1.20E+06 5.49E+07 7.79E+01 9.512+07'1.10E+06

<-Nuclide Concentrations (Ci)6.812+06 2.209+08 4.18E+07 5..49E+07 1.90E+08 1.8.0E+08 1.06E+08 1.52E+08 2.20E+08 2.40E+08.2A00OE08 1.25E+07 3.40E+07 1.77E+06 1.25E+07 5,00B+06 3.30E+07 1.50E+07 1.57E+08 1.20E+07 2.04E+08.1.90E+08 1.98E+08 1.67E208 1.60E+08 1.10E+08 5.49E+07 1.25E+08 1.80E+0.8:

1.80i+08 1.77E+08 1.90E+08 1.89E+08 1.70E+08 1.60E+08 7.09E+07 1.13E+04 4,31E+06 2.53E+05 1.80E+08 1.70Z+08 1.40E+08 2.10E+09 3.579+05 8.04E+04 1.02E+05 1.719+07 4.07E+05 2.20t+07 6.31E+06 1.30E+07 1.46E+07 1.40E+08 1.40E+08 1.60E+08 1.10E+08 9.72E+06 1.30E+08 1.40E+08 75$$ 3 3 3 3 3 3 3 3 3 3 33 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3333 <-Library Kind 2-Actinide 3-Fission Product 3 3 3 3 /3 3 3 3 3 3 2 2 2 3 3 3 2 2 2 2 2.3 3 3 3 3 3 3 3 3 3 33 t 56$$ fO t End C C C I Calc. No. STPNOC13-CALC-006 EN ER CO N CALCULATION SEMET Rev.Page No. Page 19 of 42 The resulls of this calculation are summarized below in Table 7-1. These MCNP input source definition.

values will be used in the Table 7-1 Binned Total Core Source Term Energy Energy Boundaries Photons/sec Group (e) _____(MeV)1 0.01-0.05 9.29E+19 2 0.05-0.1 2.93E+19 3 0.1-0.2 6.54E+19 4 0.2-0.3 4.28E+19 5 0.3-0.4 1.52E+19 6 0.4-0.6 3.58E+19 7 0.6-0.8 4.35E+19 8 0.8-1 2.66E+19 9 1-1.33 1.29E+19 10 1.33-1.66 1.65E+19 11 1.66-2 5.57E+18 12 2-2.5 5.53E+18 13 2.5-3 1.98E+18*....... A-4 -=4....: ..~.. --

15 4-5 3.48E+16 16 5-6.5 3.95E+11 17 6.5-8 1.75E+08 18 8-10 3.71E+07 19 10-11 2.01E+06 totals 3.95E+20 Cale. No. STPNOC13-CALC-006 EN. E C N CALCULATION SHEET Rev. 1 Page No. Page 20 of.42 7.2 MCNP Model Core Htomogeniization.

Because the source term is given for the entire core, the self-shielding ffom the assemblies is an important part of thedose rate response in regions above the core:.; Pafficles.

b6.rh in the lower section of the core are vely unlikely to penetrate thrluigh the ore itself'; aid imke it to the radiation monitors.:

For simplicity, the core is modeled as a 3 dimensionaL fylinder.

with distribiited spatial particle distribution.

The calculations for the.

i t Worksh.ef Cdmpositions:

of the EXCEL workbook STP.xlsx.

A density and isotopic..comiposition is with the water level above the top of the fuel. A summary of the fti or the ,core.: .omposition "ad ensity is shown below. The inputs are based on the dimensions in TaMbe 5-1 and tet coh i siti5 6:in Table 5-4.RodVolume=

i(Pe11 2A'Ra dius)2 x Acive igth = (3.i4)(0.16125 in)2 (168 in) = 13.7 in 3 Rod Mass9o 2= p x v (io,96 (O.95)(13.72 in) (2.s4 .= 2341.5 g Number of Fuel Rods (.:41.5 4 68 Assembly Massuo 2 = Ro d Mass X .i (2341.5 g)(264) 618.2 kg Clad Vom (D [D -.[0.374 in)2 (0.329 in)2 Clad x Active Length 1:14 4 * -. (168.in)4.17 in3............

Rod Masszry..p x V (6:56 g (4.17 in 3)(2.54---)

=448.7g cc:.' " .'Nunber.o0 Fuel RodS Assembly Masszy...4 = od a ; .R.ds.. (448.7 g)(264) = 118.5 kg Assembly Assembly H 2 0 Volume = -t[(Assembly Width)2 -,t(RodkRadius)?

x 264] x Active Length= [(8.404 in)2 -(3".4)(0,187 in)2 (264)](168 in) =.:.6993 in 3 S*"3cm3 Assembly MassH.o = p x 0 = (0.9982-)

(6993 in 3) (2.54-) = 114.4 kg Assembly Volume Active Length x (Assembly Width)2 -(168 in)(8.404 In)2 11865A in 3 Total Mass 1000(618.2

+ 118.5 kg 114.4) kg Density= Volume.. 11865.4 in 3 (2.4) 4.38 g/cc 11865.4 *n .* :r C Cale. No. STPNOC13-CALC-006 0 EN ER CON CALCULATION SHEET Rev. 1 Page No. Page 21 of 42 I 7.3 MCNP Model Geometry The following MCNP model geometry is based on the containment dimensions, summarized in Table 5-2.The model only focuses on the primary systems and components that provide shielding or reflection from the core to the radiation monitors.

These components include the reactor vessel, concrete in reactor pit, containment walls (reflection), and steam generators (reflection).

VISED plots of the model geometry are provided in Figure 7-2, Figure 7-3, and Figure 7-4. The MCNP surface cards with the model dimensions (cm) are shown in Figure 7-5, and the cell cards are shown in Figure 7-6 for the cases with no reactor head. A VISED plot of the model with the reactor head is shown in Figure 7-7. The surface and cell cards for the cases with the reactor head are shown in and Figure 7-8, respectively.

Areas that are not of interest are given an importance of zero (white areas) so MCNP will not track particles in locations that will nQt conrtribute tQA.he.detector response.

A summary of surfaces used in constructing this geometry is-shown in Table 7-2, including a description of macrobody dimensions.

Figure 7-2 X-Z VISED Plot of Reactor Vessel and Concrete Reactor Pit (No Head)Air cm 806.45 em Concrete Reactor Pit Reactor Vessel Homogenized Core Calc. No. STPNOC13-CALC-006 0 E N E RC ON CALCULATION SHEET Rev. 1'Prage No. Page 22 of 42 Figure 7-3 X-Z VISED Plot of Containment 3 I!6488.75 cm (212' 10" Ley ci)"420,14 (153' Level)1856.74 .o_ (73Level)1704.34 cm (68' Level)-=44 806A5 cm (38' 6.5" Level)4.6.11-472 Im ( 1" Level)0 cm (12" 1" Level)3 Steam Generators are not full height. Also, they are not on the same X-Z plane as the core shown above. They are included for visualization purposes.C Calc. No. STPNOC13-CALC-006 N CALCULATIONSHEET Rev. 1 Page No. Page 23 of 42 Figure 7-4 X-Y VISED Plot of the Containment Geometry at Radiation Monitor Level Radiation Monitor Steam Generators Radiation Monitor Calc. No. STPNOC13-CALC-006 EN ERCON CALCULATION SHEET A Page No. Page 24 of 42 Table 7-2 Suumairy ofSurfacesUsed for MCNP Models Surface Surface :,::; -;l;:. "- : " : ;. l/ ..;.'-": Type Number _ _Dimernsidpns Description RCC Xo Yo Z6:. Vx .V 'R 0 0- 426.72:

Active&Fuel Region 2 .0 .0 '0 0 "' 0- 0045. 219.71 .Reactor Pressure Vessel Inner Surface..3 0 .0_.-0" .0_"" 700.45. :244JI. -.` Reactor Presu-reVe'ssel Outer Surface -""31 0 0 -700.45. 0 "0 18.26': 5- 24421 -- .-..-"-ReactorPresUre:Vessel Head 41 0 .:0 .-'512.81 -" -0 676 -277. Concrete-V6id-fo-Prmar.Loop.:-

42 0 '0 .512:281 0 .',167.6: I4171 CncreteVoidlrfo irianry Loop 10 0 0 -700.45':

.. 07 64. i.. .24471 c o6crete waicutout-4 , ..-.. _ .11 444.71 843: 700,45 ."050 2 St mGenerator 1.12 444.71 .84.720. -.0 :

.7 Steam GeneratorInner.1 13 -444.71 "843 S .O7.45 " .-0 .0 -.20S0 .. c 20 .Steam.Generator2 14 -444.71 '.843 .720.45,.

0. ). .!2.0.% .230.:. teamGeneratorInner 2 15~1 -843.700.45:.

0 :'" 0:' .ta 2050e " 5.- "SteanG rnera 3 '.16 -444.71 -843 M70.45 -: .0 2010 230 ." Steam Generator4Inner3

17. 444.71: -843. 700.45 0. 0 2050;' 250 .-Stea.Ge.nator 18 444.71 -843:. 720.45 -. 0 0 -2010 230 steam rGenetor Inner.4 21 0 0 1694.34. 0 : 0 2600.8'ý .-2285 Containmentinner Liner Surface 22 0 0 .169434. .0 .i .-.26008 2285.951-

. .' Containmentflnner Concrete Surface 23 0 0 169434: .- .- 0 -26008 ..2377.39 -.Containment Outer concrete Surface 0 26 0.8 :-:23.-77..

.RPP -X x -Y -Y -z:- ::- -....4 -498 498 -498 498 -498 700. Concrete Surrounding RPV 8 -1250 1250 -437 437 II 806.45" 2116-45 Concrete Wall Fuel Pit Inner 1 9 -1356 1356 -543 543 700.45 2116.45 Concrete Wall Fuel Pit Outer SPH X0 Y 0 Zo R_

Calc. No. STPNOC13-CALC-006 E N E R C O N CALCuLATON SHEET Rev. 1 Page No. Page 25 of 42 Surface Surface Type Number Dimensions Description 5 0 0 0 219.71 1 Bottom of Reactor Pressure Vessel Inner 6 0 0 0 244.71 Bottom of Reactor Pressure Vessel Outer 24 0 0 4295.14 2285 Containment Dome Inner Liner Surface 25 0 0 4295.14 2285.95 Containment Dome Inner Concrete Surface 26 0 0 4295.14 2377.39 .Containment Dome Outer Concrete Surface PZ z 7 .0 Fuel Bottom 71 700.45 Top of RPV 20 Variable i Water Level 27 4295.14 Spring Line 28 1704.34 68' Level 101-110 42.672 426.72 Geometric Importance Divisions in Active Zone Calc. No. STPNOC13-CALC-006 SIe N E R CO".N CALCULATION SHEET Rev.Page No. Page 26 pf 42 I C Figure 7-5 MCNP Model Surface Cards 4.c surfaces 1 rcc 0 0 0 0 0 426.72 209.71 $'Active Fuel Region.J..

2 rcc 0 0 0 0 0 700.45 219.71 " $ Reactor Pressuree:vesse Ihner "srface 3 rcc 0 0 0 0 0 700A45 244.71.. .... $ Reacto. Pressure V.ssel outer Surface 31 rcc 0 0 700.45 0 0 18.26 244.71 .$ ReactorVessel, Head 4 rpp -498 498 -498 498 -498 700.45 $ .Con$.ncrete S.troundihg "RPV. .'41 rcc 0 0 512.81.0 0 167.64 274.71 $ Concrete Vid 42 rcc 0 0 512.81 0 0 167.64 411.712 $ concrete Void for Prima;ry" Lop.'5 eph 0 0 0 219.71 ." .$ Bottom of Reactor Pressure:Vessel 6 sph 0 0 0 244.71 $ Bottom of Reactor'Prds-ure" Ves&el1 7 pz 0. .$ Bottom of Active zone.71 pz 700.45 $ Top of RV 8 rpp -1250 1250 -437 437 806.45 2116.45 $ Concrete Weils FueiPif:Inner 9 rpp -1356 1356 -543 543 700.45 2116.45 $ Concrete Wall Fuel. Pit,'duter 10 rca 0 0 700.45 0 0 106 244.71 -Concrete Wall Cutbxut" 11 rcc 444.71 843 700.45 0 0 2050 250 $ Steam Generator.

1,',1/2: 12 rcc 444.71 843 720.45 0 0 2010 230 $i.Inner ste'ar' Generator I 13 rcc -444.71 843 700.45 0 0 2050 250 $ Ste m Gertor 2 14 rcc -444.71 843 720.45 0 0 2010 230 $ Inner Steam Gererator 2 15 rcca 444.71 -843 700.45 0 0 2050 250 :.$ Sta:m Geneqratoý 3 16 rcc -444..71 -843 720.45 0 0 2010 230 $ Inner Steam Generator 3 17 rca 444.71 -843 700.45 0 0 2050 250 $ stearm Generator

.18 rcc 444.71 -843 720.45 0 0 2010 230 $ inner Siýe4 Generator 4..20 pz 365.76 $ Water Elevation Surface.21 rcc 0 0 1694.34 0 0 2600.8 2285 $" Containment Inner Liner. Surface 22 rcc 0 0 1694.34 0 0 2600.8 2285.95 $.Containment Inner Con.crete

'Surface 23 rcc 0 0,1694.34 0 0 2600.8 2377.39 $.Containment outer Concrete Surface 24 sph 0 0 4295.14 2285 $ Containihent Dome. Inner Linet Surface 25 sph 0 0 4295.14 2285.95 $ Containment Dome Inner Concrete surface 26 eph 0 0 4295.14 2377.39 -. Containent Dome 0uter Concrete Surface-- -. -27 pz-+/-4Z9&.4--......--

---

$1spri-og int :-= : 28 pz 1704.34 $.:G68' Level 101 pz 42.672 $ Geometric Impo ttnce biviision Fuel: Zone 102 pz 85.344 $ Geometric .rpdrtence D'.iision

'ei Zone 103 pz 128.016 G." eometric.

Importance Division Fuel.Zone 104 pa 170.688 $ Geometric importance Division Fuel Zone 105 pz 213.36 .. $ Geometric .Impcraence bivision Fuel Zone 106 p 256.032 $ .Geometric Importance Division Fuel Zone 107 pz 298.704 $ Geometric Importance bivision .Fuel Zone 108 pz 341.376 $ Geometric importance Division Fuel Zone 109 pz 384.048 " Geometric.

Importance Ditision Fuel Zone 110 pz 426.72 $ Geometric Importance Division Fuel Zone" C The surface cards for the MCNP model without the reactor vessel head does not have surface 31. The other surfaces are identical.

(

Cale. No. STPNOC13-CALC-006 0 ENERCaN CALCULATION SHEET Rev.2 Page No. Page 27 of 42 Figure 7-6 MCNP Model Cell Cards (No Head)c cells 102 1 -4.57 -a -i01 102 1 -4.57 -1 101 -102 103 1 -4.57 -1 102 -103 104 1 -4.57 -1 103 -104 105 1 -4.57 -1 104 -105 106 1 -4.57 -1 105 -106 107 1 -4.57 -1 106 -107 108 1 -4.57 -1 107 -108 109 1 -4.57 -1 108 -109 110 1 -4.57 -1 109 -110 2 2 -0.9982 1 -3 #4 -20 4 4 -7.94 2 -3 7 -71 5 4 -7.94 5 7 #7 6 2 -0.9982--5

-7 61 2 -0.9982 -20 71 (-10:-8)71 3 -1.21E-03

-42 41 7 5 -2.3 6 3 -4 #71 8 5 -2.3 8 -9 10 9 4 -7.94 -11 12 28 10 0 -12 28 11 4 -7.94 -13 14 28 12 0 -14 28" 13 4 -7.94 -15 16 28 14 0 -16 28" 15 4 -7.94 -17 18 28 16 0 -18 28 20 4 -7.94 21 -22 21 5 -2.3 22 -23 imp: p=1 imp: P=2 irp:p=3 imp:p-4 imp :P=8 imp: p=1 6 imp:p=32 imp:p=64 imp :p=l28 imp:p-256 imp:p=256 imp :p-2 56 imp:P=256 imnpp=256 imp:p,7256 imp: p=256 imp:p=256 imp:p=256.

iMp:p=~256 imp: p=imp: p.=25 6 i mp .p ý0 imp; P=25 6 imp- p=O imp: p=25 6 imp:p=0 imp :p-256 imp:p=256$ Active Fuel Region$ Active Fuel Region$ Active Fuel Region$ Active Fuel Region$ Active Fuel Region$ Active Fuel Region$ Active Fuel Region$ Active Fuel Region$ Active Fuel Region$ Active Fuel Region$ Water Region$ RPV Shell$ Bottom RPV Shell$ Water Above Fuel .$ Water. Above Vessel Head$ Void for Primary Loop$ Concrete Surrounding RPV$ Concrete above RPV$ Steam Generator 1$ Inner Steam Generator 1$ Steam Generator 2$ Inner Steam Generator 2$ Steam Generator 3* Inner Steam Generator 3$ Steam Generator 4* $ Inner Steam Generator 4$ Containment Liner$ Containment Wall--2 2-4 -=7-ý94-2-44--~2*5ý27----

--23 5 -2.3 25 -26 27 imp:p=256

$ Containment Dome Concrete 24 5 -2.3 28 9 #21 #22 11 13 15 17 imp:p=256 30 3 -1.21E-03

(-24:-21:-8:-10:-2) 11 13 15 17 20 #8 #24 #2 1 imp:p=256 999 0 1 *2 #4 #5 #6 #7 #71 #8 #9 10#11 #12 #13 #14 #15 #16 #20 #21#.22 #23 #24 #30 #61 imp:p=O$ 68 foot level$ Air in Containment

$ Problem Boundary Calc. No. STPNOC13-CALC-006 QO K E N RCON .CACULATIONSHEET Rev. 1 Page No. Page 28 of 42 Figure 7-7 X-Z VISED Plot of Reactor Vessel and Concrete Reactor Pit (With Head)Reactor Heed ci C C Calc. No. STPNOC13-CALC-006 O NE R CO N CALCULATION SHEET_ _Page No. Page 29 of 42 Figure 7-8 MCNP Cell Cards (With Head)c cells 101 1 -4.57 101 imp:p=l $ Active Fuel Region 102 1 -4.57 -1 101 -102 imp:p=2 $ Active Fuel Region 103 1 -4.57 -1 102 -103 imp:p=3 $ Active Fuel Region 104 1 -4.57 -1 103 -104 imp:p=4 $ Active Fuel Region 105 1 -4.57 -1 104 -105 imp:p=8 $ Active Fuel Region 106 1 -4.57 -1 105 -106 imp:p=16 $ Active Fuel Region 107 1 -4.57 -1 106 -107 imp:p=32 $ Active Fuel Region 108 1 -4.57 -1 107 -108 imp:p=64 $ Active Fuel Region 109 1 -4.57 -. 108 -109 imp:p=128

$ Active Fuel Region 110 1 -4.57 -1 109 -110 imp:p=256

$ Active Fuel Region 2 2 -0.9982 1 -3 #4 -20 31 imp:p=256

$ Water Region 4 4 -7.94 2 -3 7 -71 imp:p=256

$ RPV Shell 5 4 -7.94 5 7 #7 imp:p=256

$ Bottom RPV Shell 6 2 -0.9982 7 imp:p=256

$ Water Above Fuel 62 6 -7.8212 -31 imp:p=256

$ Reactor Vessel Head 61 2 -0.9982 -20 71 (-10:-8) 31 imp:p=256

$ Water Above Vessel Head 71 3 -1.21E-03"-42 41 imp:p5=256

$ Void for Primary- Loop 7 5 -2.3 6 3 -4 #71 imp:p=256

$ Concrete Surrounding RPV 8 5 -2.3 8 -9 10 imp:p=256

$ Concrete above RPV 9 4 -7.94 -31 12 28 imp:p=256

$ Steam Generator 1 10 0 -12 28 imp:p=0 $ Inner Steam Generator I 11 4 -7.94 -13 14 28 imp:p=256

$ Steam Generator 2 12 0 -14 2`8 imp:p=0 $ Inner Steam Generator 2 13 4 -7.94 -15 16 28 imp:p=256

$ Steam Generator 3 14 0 -16 28 imp.:p=0 $ Inner Steam Generator 3 15 4 -7.94 -17 18 28 imp:p=256

$ Steam Generator 4 16 0 -18 2:8 imp:p=0 $ Inner Steam Generator 4 20 4 -7.94 21 -22 imp:p=256

$ Containment Liner 21 5 -2.3 22 -23 imp:p=256

$ Containment Wall 22 4 -7.94 24 -25 27 imp:p=256

$ Containment Dome Liner 23 5 -2.3 25 -26 27 imp:p=256

$ Containment Dome Concrete 24 5 -2.3 28 9 #21 #22 11 13 15 17 imp:p-256

$ 68 foot level 11 13 15 17 20 31 #8 #24 #2 1 imp:p=256

$ Air inside Containiment 999 0 1 #2 #4 #5 #6 #7 #71 #8 #9 #10#i1 #12 #13 #14 #15 #16 #20 #21#22 #23 #24 #30 #61 31 imp:p=0 S External to Problem Cale. No. STPNOC13-CALC-006 E N E RCON... CALCULATION SHEET Rev. 1 Page No. Page 30 of 42 7.4 MCNP Source Definition The core. source term is assumed to be uniformly distributed throughout the volume, and has an energy spectra based on the core inventory

[2]. Only the gamma source term is taket into0account for this evaluation.

Býecause the source term is generated immdliaiely after shutdown, the fuel gamma source term willI predominate.

Therefore the N-gamma. and. hard.wae activati6n source terms can be neglected (Assumiption 3). The source is defined o6 'the MCNP sd efcard using distributions to define the particle location'and energy.. Thei radius of the core is defined with the rad parameter, which automatically creates a uniforim distribution based on a cylindrical geometryr.

The ext and axs param6ters define the direction and distance of the. cylinder axis. These parameters c6rbinhed define the core Where the particles can be born. The er'g parameter defines the energy spectrumi of source particles and is based on the results of the ORLGEN-S caleulation discussed previously.

This distribution is a histogram.of energies represented by activities, Thi.e are" norialized by MCNP ti&.'create a.probability distribution.

The total activity is preserved in the tally'multipiter.;

The souce. definition cards are shown below in Figure.7-9. The sb card is a source biasing card, which in this case biases the particle geherafionlto the upper end of the core. This is a variance reduction technique to i mprovethe statistical certainty in the results.Figure 7-9 MCNP Soiurce DAfmitton Cards sdef rad=dl extc'd2 axs"0 0 1 arg=d8 <-Source Definition Card-Radius = dl-Extent = d2-Axis =+Z-Energy = d8 sii 209.72 '-Core Radius Distribution si2 h 0 42.672 85.344 128.016 170.688 213.36 256.03ý 298A04 <-Core Axial Distribution

.3 41 --376-3"84 ,'048-42"6T/2'

-....... " ... .sp2 0 i 1 1 1 1 1 1 1 1 .-Actual Uni'form Distribution sb2 0 0.001 0.001 0.01 0.0.1 0.01 0.1 0,1 0.1 1 1 .-Bia ed to Top Distribution c 'Fuel Gamma Spectra si8 h l.000e-002"5.000e-002 1.000e-001 2.000e-001 3.000e-001 4.000e-001

."(-Source Energy Groups 6.000e-601 8.000e-001 1.000e+000 1.330e+000 1.660e+000 2.000e+000*

2.500e+000 3.000e+000 4.000e+000 5.000e+000 6.500e+000 8.000e+000 1.000e+001 1.100e+.01 sp8 0.OOE+00 9..288E+19 2.926E+19 6.537E+19 4.277E+19 1.5212+19 3.578E+19

<-source Emission on Energy Basis 4.352E+19 2.66E+19 1.289E+19 1.649E+19 5.572E+18 5.527P+18 1.984E+18 7.812E+17 3.48E+163.947E+1l 1.75E+08 37100000 2009000 CI C C Cate. No. STPNOC13-CALC-006 EN ER CON CALCULATION SHET Rev. 1 PageNo.erPPJaL

&WY 31_of_ 4 I 'I IPage No. Page 31 of 42 7.5 MCNP Tally Specification The tallies used in this evaluation are point detectors placed at approximate locations of radiation monitors RE-8055 and RE-8099. Point detectors are chosen because they use quasi-deterministic dose calculations that will provide better results than surface or cell based tallies that require the particles to enter those regions. The inputs to this card are the coordinates of the dose points followed by an exclusion*zone (reduce variance), as well as a multiplier card, which represents the total core activity in photons/sec.

The tally cards are shown in Figure 7-10.Figure 7-10 MCNP Tally Cards f5c RE-8055 and RE-8099 <-Tally Comment Card f5:p -1200 -400 1909.24 20 <-Tally 5 (point detector)1200 400 1909.24 20 x y z.z exclusion-1200 -400 1909.24 20 1200 400 1909.24 20 fm5 3.947E+20

<- Tally Multiplier (Total Activity)In addition, the flux is multiplied by ANSI/ANS flux-dose conversion factors [11]. This is specified in MCNP using the de/df cards. These are shown in Figure 7-11.Figure 7-11 ANSI/ANS-6.1.1-1977 Gamma Flux to Dose Conversion Factors c --------------------------------------------------------

o ANSI/ANS-6.1.1-1977 C Gamma Flux to Dose Conversion Factors c (mrem/hr)

/ (photons/cm2-s)

C ------------------------------------------------------------------

._,_- de_-,0-1=03-05

-0.7_10ý-.

1-5.--20 --=2-30.-38-.-40--"---Ener r uI -..45 .50 .55 .60 .65 .70 .80 1. 1.4 1.8 2.2 to Dose Conversion 2.6 2.8 3.25 3.75 4.25 4.75 5. 5.25 5.75 6.25 6.75 7.5 9. 11.df0 3.96B-03 5.82E-04 2.90E-04 2.58E-04 2.83F-04 3.79s-04 (-Energy Dependent 5.01E-04 6.31E-04 7.59E-04 8.78B-04 9.85E-04 1.08E-03 Flux Multipliers 1.17E-03 1.27E-03 1.36E-03 1.44E-03 1.52E-03 1.68E-03 1.98E-03 2.51E-03 2.99E-03 3.42E-03 3.82E-03 4.01F-03 4.41E-03 4.83E-03 5.23E-03 5.60E-03 5.80E-03 6.01E-03 6.37E-03 6.74E-03 7.11E-03 7.66E-03 8.77E-'03 1.03E-02 Calc. No. STPNOC13-CALC-006 ENERCON CALCULATION SHEET ReV." .8~--e ,Page N Page_32 of 42..": " ... ... ... ... ... ,.:Page No. Page 32 of 42~1'7.6 MCNP Material Cards " " The MCNP material cards are provided in Figure 7-12...The.

are.based onthe compin bed in Table 5-4.Figuire 7-12 MCNP Material Cards. : ml 92235 -0.0245 92238 -0.5891 8016 -0.2521 40000 -0.1118 50000 -0.0017 24000 -0 .0001 26000 -0.0002 i001 -0.0211 6012. -0.ool m21001 2 8016 1 $Water m3 6012 -0.000126

$ Air 701 4 -0.76508:.

8016:-0.234793 m4 6000 -0.0008 $ Ss 304 14000 -0.01 15031 -0:00045 24000 -0.19 25055 -0.02 26000 -0.68375 ......28000 -0.095 m5. 260.00 -0.014 $ Reg-Concrete 100i -0.01 13027 -0.034.20000 -0.044 8016 -0.532 S 14000l-0.337 m6 6012 -0.01 $ Carbon Steel"26056 -0.99 C 5 Material I composition will change based on the water level relative to the core. This only applies to water heights below 14 feet.C' Calc. No. STPNOC13-CALC-006 EN.ERCO N CALCULATIONSHEET Rev.1 ptoJg E<' ft Page No. Page 33 of 42 7.7 Results File Naming Scheme: The MCNP input files are named with the following convention:

P-height-condition where: P = Project (STP)Height = water height from bottom of core (ft)Condition h -with head" --n -n nohead. .. .........

Calc. No. STPNOC13-CALC-006 0 E N ER C N CALCULATION SHEE T Rev.,Page No. Page 34 of 42 7.7.1 Results without Head The dose rate as a function of water level is provided in Table 7-3 and plotted in Figure 7-13, below.Because the MCNP model geometry is symmetric in the x and y planes, the two point detector locations should provide the same dose rate. To increase the. statistical.

certainty in the fhinal result the two individual dose rate responses and uncertainties are conbinedi.using inversevariance averaging.

All of the water levels described in the following sections refer: to the level at the top of the fuel (i.e. 0 foot water level is at the top of the fuelassemblies and '-13 feet is flange level).Table 7-3 Dose Rate Response as a Function of Water Level for no Head Configuration (rnrem/h)Water Level (ft) Dose Rate I fsd -Dose Rate 2 fsd. Dose Rate Avg Avg fsd S0 .. '9.27E+06 0.0081 9.34E+06 0.0109 § .30E+06 0,0065 2 .26E+0.5 9.0078 __ 43EO 0.0093. 4 E5 0.0060 4 2.31+04 '0.0236 ..2.32E04 0.0247 " .32E+04. 0.0171 6 1.73E+03 .3.10E-02 1.69E403 2:44E-02 1,70E+03 0.0192 8 1.51E+02:1' 0.03.02 1.51E+02 .0,87 ,. 1.51E+02 0.0208 10. 1.40E+01 " 0.036 1.36E+01 0.0323 1.38E+01 0.0240 C 0 A Calc. No. STPNOC13-CALC-006 I E N E R"C 0 N CALCULATION SHEET IRev. I Page No. Page 35 of 42 Figure 7-13 Dose Rate versus Water Height Plot for no Head Configuration 1.OOE+0 7 ----1.00E+05 Q 1.OOE+04 ___I.OE+02-1.00E+02 -' -___

-' -'--'- -__ _1.00E+-O-0 1 2 3 4 5 6 7 8 9 10 11 Water Levelfrom Top of Ibel ft)I Cale. No. STPNOC13-CALC-006 E NE R C: N CALCULATIONS ET Rev. S-'E-,j yd __________"__Rev.

Page No. Page 36 of 42 7.7.2 Results with Head.: The dose rate results for the.cases with.the head in place are the same. except the minimum detectable dose rate is lower due to the lower ainbient dose rate in the containment.

The dose rates are listed in Tableand plotted in Figure 7-14.Table.7-4 Dose Rate Response as a Function of Water LeVel for Head on Configuration.(mrem/h)

(.I I Water Level (ft). Dose Rate 1 fsd. Dose Rate 2 fsd Dose Rate Avg Avg fsd 0 2.16E+04 0.094' 0.185 2.24E+04.

0.0838 2. .1.87E4-03..

0.083 .183E+03 0.074 1.85E+03 .0.0554 4 1.11E+02 0.06i: 1.08E+'02:

0.069 1.1OE+02 0,0455 6 8.89E+"00 " .".0.085.

7.48E+00:" 0.. 0 048 7.82E+00..

0.0418 8 .95E01 0.1.25" .."1E1 ."093 8- 8.42E.-01.

0.0742.1 I-C C Calc. No. STPNOC13-CALC-006 1EN ER CO N CALCULATION SHEET Rev. 1 Page No. Page 37 of 42 Figure 7-14 Dose Rate versus Water Height Plot for with Head Configuration 1.00E+08 1,00E404 I.OOE+O -0 2 2 3 4 5 6 7 8 9 Water tevelfrom Top of Fuel (ft)

Cale. No.E.N ERCC ONN STPNOC13-CALC-006 CALCULATION SHEET Rev.I I ~Rev I Page No. Page 38 of 42 PageoNo.Page 38 of 42 Appendix A.- ENERCON Reference EMAILS (

Cale. No. STPNOCI 3-CALC-006 ENERCON CALCULATION SHEET Rev. 1 Page No. Page 39 of 42 Drew tJlakiwul!

• fwni Paxi Sudrvik Seut MondW, Pecnvmb9e.r 213 5.5 AhMuavin Moul; Jeff Gmszl Jay Tra~nar TM Cha jtweda Subject IRE. WI Rattsdng Cafty L~e'ale%we, let m atdtbn l a,:drawfota ef"'jThe watsnrlove.duui, Is the same watermfevel as the.is ....l Ow.W : t :3 -3erroo siut.I, iminpsda uekw2eise 7nelvl e3'?,n 8MR nadWftatjn"Vmbors MEM8L5.ibn~d 0114M99) iead fonsflm t m(htdurigre~palnz lithe upper-ntaane rtefrrbe&

tE wopiolynestti-C rQVWuAtbtý&

letreve! hartow, Whein thei bead iswuln daeaeuinein awttmte~dIe'ep'afarmfWfl) dmas raeaoilrrm re-ad 7 wa0ver.mjraw. The& te are~i'dse rates fras core ml Ta~ns uqyiess ihias,iWmjhrr~tis hr r ~ ffe eh-....- _ -.. .... ....... .. ..... .0..: or Nibgi lIES =ormaid oat urtleatkn prututucltbs fe rates at the monitors ý&atfaege level enU tilVý less than.SrmAN~i.

E -. ..1flO6Ttnm des0)uev TemreTeakedol Ocice: t V am 3) SS&-t kall: FaU3.I-TFN.Aneflwt l 1&zea l o o aviy*lv..'.

....A Calc. No. STPNOC13-CALC-006 D EN RC CON CALCULATION SHEET Rv Page No. Page 40-of 42 wkitw eT Q k 4 o. _AumeuatmE.0taw

... .... ..Mq&. .k b*j'Z2 4 i ad I ~ ~ ~ ~ ~ ~ ..ue~ ... .~~~so ...vnid~Inis P" 10i4i3 &6 S:b Md-~ & a~5~k, lb""..li .n wh .~su~a .a 864 MIF -feirfxnS amhhntbiii~~6~a Iaz ao16 Fa~ Nis.........

.........m ihfh#ng ~~f 73E a A b~et~f rn...... .. .bR~cukt~,n.............~

d'~the (1.. ..C C Catc. No. STPNOC13-CALC-006 P NNERPCa oN CALCULATION SHET Rev.Page No. Page 41 of 42 Appendix B -Electronic File Listing Volume in drive F is My Passport Volume Serial Number is 1AEA-6007 Directory of F:\STPNOC013-CALC-006\Rev 1 03/14/2014 04:12 PM <DIR>03/14/2014 03/21/2014 02/06/2014 02/07/2014 03/14/2014 03/21/2014 03/14/2014 02/07/2014 03/14/2014 03/14/2014 03/14/2014 03/14/2014 04;12 0.9:33 02:03 10:26 08:44 09:32 04:12 12:14 09:10 08:48 04:06 04:02 PM PM PM AM AM PM PM PM AM AM PM PM<DIR><DIR><DIR>0 dir.dat 100,953 EMAIL from Paul Sundak, Dec. 9 2013.pdf 8,795 Inverse Variance Weighting.xlsx 332,025 liner plate info.pdf mcnp origen 111,247 RE Fuel Assembly Dimension.pdf 462,166 RPV with core.pdf 537,808 RPV.pdf 43,842 STP.xlsx 1,036,800 STPNOC013-CALC-006 Rl.doc 2,633,636 bytes 9 File(s)Directory of F:\STPNOC013-CALC-006\Rev l\mcnp 03/21/2014 03/21/2014 03/21/2014 03/21/2014 02/0.6/2014 03/21/2014 03/21/2014 03/21/2014 03/21/2014 09:32 09:32 09:32 09:32 11:31 09: 34 09:45 09:45 09:45 PM PM PM PM AM AM AM AM AM<DIR><DIR><DIR><DIR>head no head 137 STP.bat 18,720 STP.sx 4,053 STP default.sx 9,744 _sx.log 2,007 _sx.var 34,661 bytes 5 File(s)Directory of F:\STPNOC013-CALC-006\Rev l\mcnp\head

-- .------~-0-3/-2~1/-20-14~09 32~PM-"~ < D IR ~ '-~ ~ ~ -_________

___________

... -----.. I 03/21/2014 03/12/2014 03/12/2014 03/12/2014 03/21/2014 03/21/2014 03/21/2014 03/21/2014 03/21/2014 03/14/2014 03/14/2014 03/14/2014 03/21/2014 03/21/2014 03/21/2014 03/21/2014 03/21/2014 03113/2014 03/13'/2014 03/13/2014 03/21/2014 03/21/2014 03/21/2014 03/21/2014 03/21/2014 03/12/2014 03/12/2014 03/12/2014 09:32 PM 12:51 PM 08:50 PM 08:50 PM 09:45 AM 04:41 PM 09:17 PM 04:41 PM 09:17 PM 08:27 AM 03:42 PM 03:42 PM 09:45 AM 04:43 PM 09:10 PM 04:43 PM 09:10 PM 04:35 PM 08:40 PM 08:40 PM 09:45 AM 04.43 PM 09:17 PM 04:43. PM 09:17 PM 01:17 PM 08:51 PM 08:51 PM<DIR>8,990 STP14h5 1,104 STPl4h5m 924,317 STP14h5o 8,587 STP14h8 1,260 STPl4h8m 1,312 STPI4h8m2 545,780 STPl4h8o 557,996 STP14h8o2 8,990 STPI6h7 1,260 STP16h7m 942,029 STP16h7o 8,587 STP16h8 1,312 STPl6h8m 1,364 STPl6h8m2 557,572 STPI6h8o 543,468 STPI6h8o2 8,990 STPI8h6 1,156 STPl8h6m 552,616 STP18h6o 8,587 STPl8h8 1,260 STP18h8m 1,312 STPl8h8m2 551,487 STPI8h8o 565,735 STPl8h8o2 8,989 STP20hS 1,104 STP20h5m 966,684 STP20hSo Calc. No. STPNOC13-CALC-006 O ENE RIC 0 N CALCULATION SHEET Rev.Page No. Page 42 of 42 03/21/24014 09:45 AM 8,586 ST?20h8 03/21/2014 01:52 PM 1,260 STP20h8m 03/21/2014 07:40 PM 1,364 STP20h8m2 03/21/2014 01:52 PM 550,ý35 STP20h8o 03/21/2014 07:40 PM 658,209 STP2Oh8o2 03/12/2014 01:17 PM 6,990 STP22h5 03/12/2014 08:51 PM 1,104 STP22h5m 03/12/2014 08:51 PM 936,997 STP22h5o 03/21/2014 09:45 AM 8,587 STP22hS 03/21/2014 01:52 PM 1,260 STP22h8m 03/21/2014 07:07 PM ..1,364.STP22h8m2 03/21/2014 01:52 PM 547,911.STP22h8o 03/21/2014 07:07 PM 625,528 STP22h8o2 40 Filels) 10,133,743 bytes Directory of F:\STPNOC013-CALC-006\Rev 1\mcnp\no hea-d-03/21/2014 09:32 PM <DIR>03/21/2014 09:32 PM <DIR) >.03/13/2014 04:57 PM 8,270 STiP14n7..03/14/20.14.

09:14 AM .-1,364 SuP14zj7mý 03/14/2014 09:14 AM .782,051 STP14n7o 03/21/2014 09:45 AM 8,468 sTP14n*B 03/21/2014 03:46 PM 1,364 STP148'm 03/21/2014 03:46 PM 447,366 STM14n8o 03/13/2014 04:57 PM 8,272 ST016n7 03/14/2014 11:49 AM 1,364 STP16D7m 03/14/2014 11:49 AM 702,157'STP16n7o 03/21/2014 09:45 AM 8,468* STPI6n8 03/21/2014.

03:46 PM 1,416 STP16nm .03/21/2014 03:46 PM 437,587 STP16n8o 03/13/2014 04:35 PM 8,863 STP!8n6 03/14/2014 01:10 AM 1,520 STPl6n6m .'03/14/2014 01:i0 AM 541,013 STP18n6o 03/21/2014 09:45 AM 8,468 STPln8.03/21/2014 02:02 PM 1,572 STPISn8m 03/2j/2014 02:02 PM 534,616 STP18n8o.-. .....8.268STP20.n7 L---_ " 03/14/2014 12:32 PM.. i364 STP20n7m 03/14/2014 12:32 PM 902,858 STP2On7o 03/21/2014 09:45 AM .5,467 STP20n8 03/21/2014 03:'46 PM 1,312 STP20n8m 03/21/2014 03;46 PM 460,778 STP20n8o 03/13/2014 04:56 PM 8,273 STP22n7 03/14/2014 08:58 AM. 1,364 8TP22n7m 03/14/2014 08:58 AM 888,128 STP22n7o 03/21/2014 09':45 AM 8.,468 STP22n8.03/21/2014 03:46 PM 1,312 STP22n8m 03/21/2014 03:46 PM 464,349 STP22n~o 03/13/2014 04:56 PM 8,272 STP24n7'03/14/2014 12:45 PM 1,364 STP24n7m 03/14/2014 12:45 PM 870,869 8TP24n7o 03/21/20.14 09:45 AM 8,468 STP24n8 03/21/2014 03:46 PM 1,312 S.TP24n8m' 03/21/2014 03:46 PM 450,196 8TP24n8o 36 File(s) 7,599,621 bytes Directory of F:\STPNOC013-CALC-006\Rev 1\origen 03/14/2014 04:12 PM <DIR>03/14/2014 04:12 PM <DIR>02/03/2014 07:26 PM 2,416 STPEAL.inp 02/04/2014 02:43 PM 99,996 STPEAL.out 2 File(s) 102,412 bytes A IL C1 C OPOP03-ZG-0009 Rev. 59 Page 59 of 115 Mid-Loop Operation Addendum 1 RCS/RHR Simplified Elevation Diagram Page 1 of 1 REACTOR COOLANT SYSTEM PRESSURIZER SECTION A-A HOT LEG STP D-0794 Rev 2 0POP03-ZG-0009 Rev. 59 Page 60 of 115 Mid-Loop Operation Addendum 2 RVWL Sensor Elevations Page 1 of I NOTE" Top of Core is elevation 28 ft 2 inches." SG spillover is elevation 34 ft 3.8 inches.SENSOR UPPER HEAD PLENUM INDICATED INDICATED SENSOR LEVEL DESCRIPTION LEVEL (%) LEVEL (%)All Covered 100 100 46' 4.75" Upper Head Full 1 64 100 45' 3.4" Upper Head Partially Drained 2 0 100 39' 4.9" Plenum Full 3 0 85 34' 11 Plenum NOT Full (Enter Reduced Inventory) 4 0 66 33' 5.5" Top of Hot Leg Nozzle 5 0 48 32' 3" Hot Leg Centerline 6 0 33 31'0.5" Bottom of Hot LegNozzle 7 0 20 30' 1.6" Midway between Hot Leg Nozzle and Upper Core Plate 8 0 0 29' 2.7" Upper Core Plate

STPEGS UFSAR TABLE 12.3.4-1 AREA RADIATION MONITORS Reactor Containment Building Tag Number and Location Range (mR/hr) (3) High Alarm Setpoint (mr/hr) (2)NIRA-RE-8052 10-1_10 4 1,000 Incore Instrumentation Room (- 1 ft-6 in.)NIRA-RE-8053 10--1_0 4 100 Support across from elevator (-11 ft-3 in.)N1RA-RE-8054 I0"-10 4 100 West Stair Landing (19 ft-0 in.)N1RA-RE-8055 101_104 100 North SG wall across from the head laydown area (68 ft-0 in.)NI RA-RE-8056 10"-104 100 Support across from elevator (52 ft-0 in.)NI RA-RE-8099 101-10 4 100 South SG wall across from the in-containment fuel pool (68 ft-0 in.)1. Tag Number is shown for Unit 1. For Unit 2, change second position from 1 to 2.2. The alarm setpoints listed are typical and may be varied as necessary.

3. Range is based on microprocessor conversion factor and a detector signal which has a high degree of confidence.

Conversion factor will vary dependent on the detector calibration.

Exact ranges are found in plant instrument scaling manuals.12.3-24 Revision 15 STPEGS UFSAR TABLE 12.3.4-1 (Continued)

AREA RADIATION MONITORS Fuel Handling Building Tag Number and Location (1) Range (mR/hr) High Alarm Setpoint (mr/hr) (2)N1RA-RE-8081 10-1_104 2.5-11 ft S ofcols. 30.2 and $5 (68 ft-0 in.)NIRA-RE-8084 10-1_104 2.5-24 ft S of cols. 28 and T 5 (-21 ft-0 in.)NIRA-RE-8085 10-1_04 2.5-24 ft S of col. 28 and-6 ft E ofcol. S 5 (-21 ft-0 in.)NIRA-RE-8086 10-1_104 2.5-24 ft S of col. 28 and-11 ft E of col. R, (-21 ft-0 in.)N1RA-RE-8087 10-104 2.5 col. 30.2 and 12 ft W of col. R 1 (4 ft-0 in.)N1RA-RE-8088 10 1-10 4 2.5 3 ft S of col. 30.9 and col. R 1 (30 ft-0 in.)N1RA-RE-8089 10"-1_0 4 2.5 col. 28 and col. N (68 ft-0 in.)NIRA-RE-8090 10"1_104 2.5 18 ft N of col. 30.2 and col. T 5 (68 ft-0 in.)NIRA-RE-8091 10-1_104 2.5 col. 34 and col. N (68 ft-0 in.)1. Tag Number is shown for Unit 1. For Unit 2, change second position from 1 to 2.2. The alarm setpoints listed are typical and may be varied as necessary.

12.3-25 Revision 15 STPEGS UFSAR TABLE 12.3.4-1 (Continued)

AREA RADIATION MONITORS Tag Number and Location (NIRA-RE-8097 33 ft S of cols. 28 and 10 ft W of col. N (68 ft-0 in.)Fuel Handling Building (Continued)

Range (mR/hr)10-2-107 High Alarm Setpoint (mr/hr) (2)1,000 1. Tag Number is shown for Unit 1. For Unit 2, change second position from 1 to 2.2. The alarm setpoints listed are typical and may be varied as necessary 12.3-26 Revision 15 STPEGS UFSAR TABLE 12.3.4-1 (Continued)

AREA RADIATION MONITORS Mechanical Electrical Auxiliaries Building Tag Number and Location 1 Range (mR/hr) (3) High Alarm Setpoint (mr/hr) (2)NIRA-RE-8057 102-210 0.5 col. 22 and -10 ft E of col. J (10 ft-0 in.)NIRA-RE-8058 10-1i0 4 2.5 col. 26 and col. J (10 ft-0 in.)NIRA-RE-8059 101-104 2.5 col. 27 and col G (10 ft-0 in.)NIRA-RE-8060 10-1-104 2.5-10 ft S of col. 30 and col. E (10 ft-0 in.)N1RA-RE-8061 10"1-10 4 2.5-10 ft S of col. 24 and-11 ft W of col. E (10 ft-0 in.)NI RA-RE-8062 10-1-10 4 2.5-6 ft S of col. 31 and col. C (10 ft-0 in.)NIRA-RE-8063 I0-1_0 4 2.5-9 ft S of col. 28 and col. B (10 ft-0 in.)NI RA-RE-8064 101-1_0 4 2.5-12 ft S of col. 24 and col. F (29 ft-0 in.)N1RA-RE-8065 10-1_104 2.5-5 ft N of col. 32 and col. C (29 ft-0 in.)I. Tag Number is shown for Unit 1. For Unit 2, change second position from 1 to 2.2. The alarm setpoints listed are typical and may be varied as necessary.

3. Range is based on microprocessor conversion factor and a detector signal which has a high degree of confidence.

Conversion factor will vary dependent on the detector calibration.

Exact ranges are found in plant instrument scaling manuals 12.3-27 Revision 15 STPEGS UFSAR TABLE 12.3.4-1 (Continued)

AREA RADIATION MONITORS Mechanical Electrical Auxiliaries Buildinn (Continued)

Tag Number and Location ( Range (mR/hr) (3) High Alarm Setpoint (mr/hr) (2)NI RA-RE-8066 10-2-103 0.5-4 ft N of col. 22 and 14 ft E of col. C (35 ft-0 in.)NI RA-RE-8067 10"-2103 0.5 col. 22 and 10 ft E of col. J (35 ft-0 in.)NI RA-RE-8068 10"1-10 4 2.5-10 ftN of col. 25 and col. H (41 ft-0 in.)N2RA-RE-8068 I0"1-104 2.5-10 ft S of col. 24 and col. G (41 ft-0 in.)NI RA-RE-8069 102_10 3 0.5-12 ft S of col. 24 and-14 ft E of col. C (41 ft-0 in.)NiRA-RE-8070 10-2_103 2.5 col. 29 and col. C (41 ft-0 in.)NI RA-RE-8071 10-2_103 2.5-18 ft S of col. 28 and 3 ft W of col. B (41 ft-0 in.)NI RA-RE-8072 10-1_104 100-11 ftN of col. 29 and 5 ft W of col. D (41 ft-0 in.)NIRA-RE-8073 10-1_104 2.5 col. 29 and col. E (41 ft-0 in.)NIRA-RE-8074 10-10 4 2.5-5 ft S of col. 31 and-7 ft W of col. C (41 ft-0 in.)1. Tag Number is shown for Unit 1. For Unit 2, change second position from 1 to 2.2. The alarm setpoints listed are typical and may be varied as necessary.

12.3-28 Revision 15 STPEGS UFSAR TABLE 12.3.4-1 (Continued)

AREA RADIATION MONITORS Mechanical Electrical Auxiliaries Building (Continued)

Tag Number and Location (1) Range (mR/hr) High Alarm Setpoint (mr/hr) (2)NiRA-RE-8075 10-1-10 4 15.0 col. 28 and -3 ft W of col. G (41 ft-0 in.)N1RA-RE-8076

-102_1013 0.5 col. 22 and -10 ft E of col. J (60 ft-0 in.)NiRA-RE-8077 10-1_10 4 2.5 col. 27 and col. J (60 ft-0 in.)N1RA-RE-8078 101-1_0 4 15.0 col. 27 and col. F (60 ft-0 in.)NI RA-RE-8079 10-1_10 4 15.0 col. 25 and -2 ft W of col. F (60 ft-0 in.)NlRA-RE-8080 10-I10 4 2.5 col. 26 and col. H (41 ft-0 in.)N1RA-RE-8082 101-1_04 2.5 col. 28 and -8 ft E of col. H (69 ft-0 in.)N1RA-RE-8083 10-1_10 4 15.0-10 ft S of col. 29 and 15 ft W of col. E (41 ft-0 in.)NI RA-RE-8098 102_107 1000-6 ft N of col. 25 and col. H (60 ft-0 in.)1. Tag Number is shown for Unit 1. For Unit 2, change second position from 1 to 2.2. The alarm setpoints listed are typical and may be varied as necessary.

12.3-29 Revision 15 STPEGS UFSAR TABLE 12.3.4-1 (Continued)

AREA RADIATION MONITORS Tag Number and Location ()NIRA-RE-8092 col. 9 and col. P TGB (29 fi-0 in.)NIRA-RE-8093 col. 7 and col. M TGB (29 ft-0 in.)NIRA-RE-8094

-3 ft N of col. 23 and-14 ft W of col. B TSC-MEAB (72 ft-0 in.)Miscellaneous Buildings Range (mR/hr)10-2_103 10-2_103 10-2_107 High Alarm Setpoint (mr/hr) (2)0.5 0.5 1000 z 1. Tag Number is shown for Unit 1. For Unit 2, change second position from I to 2.2. The alarm setpoints listed are typical and may be varied as necessary.

12.3-30 Revision 15 STPEGS UFSAR TABLE 12.3.4-1 (Continued)

AREA RADIATION MONITORS Tag Number and Location (1)A1RA-RE-8050 RCB (68 ft-0 in.)CIRA-RE-8051 RCB (68 ft-0 in.)Post-Accident Monitors Range (R/hr)100-108 100_108 High Alarm Setpoint (R/hr) (2)2000 2000 1. Tag Number is shown for Unit 1. For Unit 2, change second position from 1 to 2.2. The alarm setpoints listed are typical and may be varied as necessary.

12.3-31 Revision 15 REV. 21 OPOP05-EO-EO10 LOSS OF REACTOR OR SECONDARY COOLANT PAGE 13 OF 27 S-P ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED I CAUTION Containment H2 concentration should be continuously monitored following a LOCA. to avoid explosive H2 concentrations.

I 12 MONITOR Containment H2 Concentration:

a. Containment H2 -GREATER THAN OR EQUAL TO ZERO (QDPS QUAL PAMS)b. H2 concentration

-GREATER THAN OR EQUAL TO 0.5%a. PLACE containment H2 monitoring system in service per ADDENDUM 1, ESTABLISHING CONTAINMENT H2 MONITORING.

b. PERFORM the following:

1) WHEN H2 concentration is GREATER THAN 0.5%, THEN PERFORM Step 12.c and 12.d.2) GO TO Step 13.c. H2 concentration

-LESS THAN 4% BY VOLUME c. PERFORM the following:

1) CONSULT TSC staff for additional recovery actions.2) GO TO Step 13.d. PLACE hydrogen recombiners in service per 0POP02-CG-0001, ELECTRIC HYDROGEN RECOMBINERS OPOP04-RC-0003 Excessive RCS Leakage Rev. 18 Page 53 of 127 endum ass ass Pa of7 STEP DESCRIPTION FOR OPOP04-RC-0003 STEP 3.0 STEP: CHECK Trends For Any Of The Following Indications Of RCS Leakage:* Rad Monitor RT801 1 Particulate

-Rising* Reactor Coolant Drain Tank Level -Rising* Pressurizer Relief Tank Level -Rising* RCB Normal Sump Level -Rising PURPOSE: To determine if leakage is from RCS and not CVCS.BASIS: Indication of RT801 1, RCDT, PRT or RCB Normal Sump levels rising will confirm that the leakage is from RCS and not CVCS which is normally tied to the RCS.ACTIONS: Monitor trends from RT80 11, RCDT, PRT or RCB Normal Sump.INSTRUMENTATION:

Level indications located on CP004 and various plant computer monitors located in control room. Radiation Monitor Computer RM-1 1.CONTROL/EQUIPMENT:

N/A KNOWLEDGE:

N/A This Procedure is Applicable in Modes 1, 2, 3, and 4 EHU1 NRC FORM 651 U.S. NUCLEAR REGULATORY COMMISSION (10-C0042 CERTIFICATE OF COMPLIANCE FOR SPENT FUEL STORAGE CASKS Page 1 of 4 711-.a U.S. Nuclear Regulatory Commission is issuing this Certificate of Compliance pursuant to Title 10 of the Code of Federal Regulations, Part 72, "Licensing Requirements for Independent Storage of Spent Nuclear Fuel and High-Level Radioactive Waste" (10 CFR Part 72). This certificate is issued in accordance with 10 CFR 72.238, certifying that the storage design and contents described below meet the applicable safety standards set forth in 10 CFR Part 72, Subpart L, and on the basis of the Final Safety Analysis Report (FSAR) of the cask design. This certificate is conditional upon fulfilling the requirements of 10 CFR Part 72, as applicable, and the conditions specified below.Certificate No. Effective Expiration Date Docket No. Amendment No. Amendment Effective Date Package Identification No Date 1032 June 13, June 12, 72-1032 0 USA/72-1032 2011 2051.I Issued To: (Na Holtec IntE Holtec Ce 555 Linco Marlton, N Safety Analysis Holtec .1 Final Sa HI-STOI This

• endix A S,.cified be APPROVEN Model No.: 'El me/Address) ernational nter In Drive West JJ 08053 sReport Title nternational fety Analysis Repo r9)the RM FW MPC Stora'gSystem ate is conditioned 4 mn fulfilling t te C as applic bp, the attached (Technical Speci '51ions) and A x pove CAo t Design Fe tures), and the conditions low: HI-STORM FW M DESCRIPTION:

" The HI-STORM FW M fora!!!System che follo ing corn ents: (1) interchangeable multi-purpose canisters (MwPiC51h contain the f&(2) a storage ovrae (HI-STORM FW), which contains the MPC during storage; and (3)e¢tansfer cask (HI-TRAC.VW), whichi ontains the MPC during loading, unloading and-transfer-operations T-he-MIARC r-p-to-37-pre~su ru-dwater-reactor-fu el-assem blies-or-upto-89-bciling.

water reactor fuel assemblies. -" The HI-STORM FW MPC Storage. System is certified as described in-the Final Safety Analysis Report (FSAR)and in the U. S. Nuclear Regulatory Commission's (NRC) Safety Evaluation Report (SER) accompanying the Certificate of Compliance (CoC).The MPC is the confinement system for the stored fuel. It is a welded, cylindrical canister with a honeycombed fuel basket, a baseplate, a lid, a closure ring, and the canister shell. All MPC components that may come into contact with spent fuel pool water or the ambient environment are made entirely of stainless steel or passivated aluminum/aluminum alloys. The canister shell, baseplate, lid, vent and drain port cover plates, and closure ring are the mainconfinement boundary components.

All confinement boundary components are made entirely of stainless steel. The honeycombed basket provides criticality control.

I. -NRC FORM 651 (3-1999)10 CFR 72 CERTIFICATE OF COMPLIANCE FOR SPENT FUEL STORAGE CASKS Supplemental Sheet U.S. NUCLEAR REGULATORY COMMISSION Certificate No. 1032 Amendment No. 0 Page 2 of 4 DESCRIPTION (continued)

There are two types of MPCs: the MPC-37 and MPC-89. The number suffix indicates the maximum number of fuel assemblies permitted to be loaded in the MPC. Both MPC models have the same external diameter.The HI-TRAC VW transfer cask provides shielding and structural protection of the MPC during loading, unloading, and movement of the MPC from the cask loading area to the storage overpack.

The transfer cask is a multi-walled (carbon steel/lead/carbon steel) cylindrical vessel with a neutron shield jacket attached to the exterior and a retractable bottom lid used during transfer operations.

The HI-STORM FW storage overpack provides shielding and structural protection of the MPC during storage.The overpack is a heavy-walled steel and concrete, cylindrical vessel. Its side wall consists of plain (un-reinforced) concrete that is enclosed between inner and outer carbon steel shells. The overpack has air inlets at the bottom and air outlets at the top to allow air to circulate naturally through the cavity to cool the stored MPC. The inner shell has supports att ched.to iourface to guide the MPC during insertion and removal and provide a meansto prq h.of nfPot0j oundary against impactive or impulsive loadings.

A loaded MPC is s~oye'd+itliin the HI-STORM FW'1obagverpack in a vertical orientation.

CONDITIONS

1. OPERATING PROCEDI.Written operating pedures* maintenance.

Thpu u-er's site described in Chapf'9 of the 2. ACCEPTANCE M 1 Written acce ptan ests a~described in Chap er 10 of tb confinement weld Iýfl 2m boundary welds le rate4 leakage rate exceec"O"6the a(the area repaired per$8JE,(be performed until the g 3. QUALITY ASSURANCE S ve mnt, surveillance, and be corl'sistent with the technical basis ponsisnt with the technical basis shell'baseplate, an MPC sss s~rometer.

The confinement A 3 14.5 to "leaktight" criterion.

If a I oi;e.kage shall be determined and"j0 requirements, Re-testing shall Activities in the -areas of design, ~ ion~ ~se> ,ins t. t ~________

_____________

-Pu ds iainks y, inspection, testing, operation, maintenance, repair, modification of structures, systems atd cr5mporfents, and decommissioning that are important-to-safety shall be conducted in accordance with a Commission-approved quality assurance program which satisfies the applicable requirements of 10 CFR Part 72, Subpart G, and which is established, maintained, and executed with regard to the storage system 4. HEAVY LOADS REQUIREMENTS Each lift of an MPC, a HI-TRAC VW transfer cask, or any HI-STORM FW overpack must be made in accordance to the existing heavy loads requirements and procedures of the licensed facility at which the lift is made. A plant-specific review of the heavy load handling procedures (under 10 CFR 50.59 or 10 CFR 72.48, as applicable) is required to show operational compliance with existing plant specific heavy loads requirements.

Lifting operations outside of structures governed by 10 CFR Part 50 must be in accordance with Section 5.2 of Appendix A.II'p I.NRC FORM 651 1I(3-1999) 1 10 CFR 72 CERTIFICATE OF COMPLIANCE FOR SPENT FUEL STORAGE CASKS Supplemental Sheet U.S. NUCLEAR REGULATORY COMMISSION Certificate No. 1032 Amendment No.0 Paqe 3 of 4 5. APPROVED CONTENTS Contents of the HI-STORM FW MPC Storage System must meet the fuel specifications given in Appendix B to this certificate.

6. DESIGN FEATURES Features or characteristics for the site or system must be in accordance with Appendix B to this certificate.

7. CHANGES TO THE CERTIFICATE OF COMPLIANCE The holder of this certificate who desires to make changes to the certificate, which includes Appendix A (Technical Specifications) and Appendix B (Approved Contents and Design Features), shall submit an application for amendment of the certificate P 8. SPECIAL REQUIREMENTS FORT YSTEMS IN PLACE The air mass flow rate thIqu.-h the cask system will be determined bylde ct measurements of air velocity in the overpack cooling passl for the first HI-STORM FW MPC Cask Systeol~laced into service by any user with a heat load equal to org e than 30 kW. The velocity will be measured 1 J irghe annulus formed between the MPC shell and the cpacd'1ln'r shell. An analysis shall be pe oetfn6 tht'c!e monstrates the measurements Chapter 4 of the FN. -I-A letter report su'!p arizing the in accordance withf;lO C FR .report submitted

_1t{he N 9. PRE-OPERATIONALCSTIN A dry run training rdrise t HI-STORM FW MPrdtoragef load spent fuel asserb1tes T1 run may be performedin.

all performed.

The dry-ru n-9j A Movina the MPC and thell ,hall be submitted to the NRC Fencing a validation test 1sfer of thethe first use of the system to Rh spent fuel in the MPC. The dry lures, but all steps must be'1K!'b. Preparation of the HI-STORM FW MPt6 Stedge 9'y~tem for fuel loading.c. Selection and verification of specific fuel assemblies to ensure type conformance.

d. Loading specific assemblies and placing assemblies into the MPC (using a dummy fuel assembly), Including appropriate independent verification.

e. Remote installation of the MPC lid and removal of the MPC and transfer cask from the spent fuel pool or cask loading pool.f. MPC welding,.NDE inspections, pressure testing, draining, moisture removal (by vacuum drying or forced helium dehydration, as applicable), and helium backfilling. (A mockup may be used for this dry-run exercise,)

g. Transfer of the MPC from the transfer cask to the overpack.K I .

1 NRC FORM 651 U.S. NUCLEAR REGULATORY COMMISSION S(31999) CERTIFICATE OF COMPLIANCE Certificate No. 1032 10CFR72 FOR SPENT FUEL STORAGE CASKS Amendment No. 0 Supplemental Sheet Page 4 of 4 h. Placement of the HI-STORM FW MPC Storage System at the ISFSI.i. HI-STORM FW MPC Storage System unloading, including flooding MPC cavity and removing MPC lid welds. (A mockup may be used for this dry-run exercise.)

Any of the above steps can be omitted if they have already been successfully carried out at a site to load a HI-STORM 10 System (USNRC Docket 72-1014).10. AUTHORIZATION The HI-STORM FW MPC Storage System, which is authorized by this certificate, is hereby approved for general use by holders of 10 CFR Part 50 licenses for nuclear reactors at reactor sites under the general license issued pursuant to 10 CFR 72.210, subject to the conditions specified by 10 CFR 72.212, this certificate, and the attached Appendice..

Pcpd I he, IISTORM FW MPC Storage System may be fabricated and used in accordance n ent to CoC No. 1032 listed in 10 CFR 72.214.Each of the licensed HI-STOM oP C Storage System (i.e., the MPC, overpack, and transfer cask), if fabricated in accqr e d with any of the approved CoC kindments, may be used with one another provided an assessmentý

~performed by the CoC holder that demonstes design compatibility.

The HI-STORM FW MPC Stopg. System may be installed on an ISFSI pad wi$:the HI-STORM 100 Cask System (USNRC Docket 72-1l'4)rovided an assessment is performed by the Cp older that demonstrates design___ -compatibility.

--p-, ." j..E N C EA IULATORYI&OMM.SS.ON PINLA In ench sn Storagedand&Transportation u eia

• n, 0555 Dated July 14, 2011 Attachments:

1. Appendix A 2. Appendix B I

....-CERTIFIGATE-OF- COMPLIANCE-NO.

1032 -- -APPENDIX A TECHNICAL SPECIFICATIONS FOR THE HI-STORM FW MPC STORAGE SYSTEM Programs 5.0 5.0 ADMINISTRATIVE CONTROLS AND PROGRAMS (continued)

5.3 Radiation

Protection Program 5.3.1 Each cask user shall ensure that the Part 50 radiation protection program appropriately addresses dry storage cask loading and unloading, as well as ISFSI operations, including transport of the loaded OVERPACK or TRANSFER CASK outside of facilities governed by 10 CFR Part 50. The radiation protection program shall include appropriate controls for direct radiation and contamination, ensuring compliance with applicable regulations, and implementing actions to maintain personnel occupational exposures As Low As Reasonably Achievable (ALARA). The actions and criteria to be included -in the program are provided below.5.3.2 As part of its evaluation pursuant to 10 CFR 72.212(b)(2)(i)(C), the licensee shall perform an analysis to confirm that the dose limits of 10 CFR 72.104(a)will be satisfied under the actual site conditions and ISFSI configuration, considering the planned number of casks to be deployed and the cask contents.5.3.3 Based on the analysis performed pursuant to Section 5.3.2, the licensee shall establish individual cask surfacedose rate limits for the TRANSFER CASK and the OVERPACK to be used at the site. Total (neutron plus gamma) dose rate limits shall be established at the following locations:

a. The top of the OVERPACK.b. The side OVERPACK c. The side of the TRANSFER CASK d. The inlet and outlet ducts on the OVERPACK 5.3.4 Notwithstanding the limits established in Section 5.3.3, the measured dose rates on a loaded OVERPACK or TRANSFER CASK shall not exceed the following values: a. 30 mrem/hr (gamma + neutron) on the top of the OVERPACK b. 300 mrem/hr (gamma + neutron) on the side of the OVERPACK, excluding inlet and outlet ducts c. 3500 mrem/hr (gamma + neutron) on the side of the TRANSFER CASK 5.3.5 The licensee shall measure the TRANSFER CASK and OVERPACK surface neutron and gamma dose rates as described in Section 5.3.8 for comparison against the limits established in Section 5.3.3 or Section 5.3.4, whichever are lower.(I.Certificate of Compliance No. 1032 Amendment No. 0 Appendix A 5.0-3 HUI OERPO1-ZV-SH03 Rev. 12 Page2 of 13 Acting Security Manager 1.0 Purpose and Scope 1.1 This procedure specifies the actions to be completed by the Acting Security Manager during a declared emergency.

1.2 This procedure implements the necessary Security emergency response actions for an Unusual Event and for initial immediate response for higher emergency classifications until relieved by the Security Manager.1.3 This procedure implements the requirements of the South Texas Project Electric Generating Station (STPEGS) Emergency Plan specific to the Acting Security Manager.2.0 Responsibilities 2.1 The Security Force Supervisor assumes the responsibilities of the Acting Security Manager until relieved.

Those responsibilities include: 2.11 -1 Directing thepimplereritati rgerY resonse activities.

2.1.2 Implementing

assembly and accountability efforts.2.1.3 Assisting with Protected and Owner Controlled Area evacuation.

2.1.4 Establishing

special access controls.2.1.5 Providing for the expedient entry/exit of emergency vehicles.2.1.6 Directing changes to security operations based on radiological conditions.

2.1.7 Determining

level of compliance with current security procedures.

3-. 0- Precautions-and--imitations..

3.1 OERPO1-ZV-IN04, Assembly and Accountability are required at a Site Area Emergency Classification or greater unless to do so would put site personnel at risk. The Emergency Director at anytime as dictated by conditions may order assembly and Accountability.

3.2 OERP01-ZV-IN05, Site Evacuation is required at a Site Area Emergency Classification or greater unless to do so would put site personnel at risk. The Emergency Director at anytime as dictated by conditions may order site Evacuation.

IOERP01-ZV-SH03 Rev. 12 Page3 of 13 Acting Security Manager I 4.0 References

4.1 STPEGS

Emergency Plan 4.2 OERPO1-ZV-IN03, Emergency Response Organization Notification 4.3 OERP01-ZV-IN04, Assembly and Accountability 4.4 OERP01-ZV-IN05, Site Evacuation

4.5 OERPO

1 -ZV-RE02, Documentation 4.6 OPOP04-ZO-0007, Aircraft Crash Onsite 4.7 OPGP05-ZV-0004, Emergency Plan hIplementing Procedure Users Guide 4.8 Security Instruction SI 2202, Owner Controlled Area Vehicle Patrol 5.0 Procedure 5.1 IF an Unusual Event or higher emergency classification is declared, implement Data Sheet 1, Acting Security Manager Checklist.

Use Checklist to help direct emergency activities.

5.2 IF contacted by the Security Manager, provide a briefing of the current situation and the security activities underway using Data Sheet 2, Security Briefing Checklist.

5.3 WHEN responsibilities have been transferred to the Security Manager, THEN return to the implementation of Security procedures and discontinue the use of this procedure.

5.4 During

an Alert or higher classification, ensure an ERO Qualified EMT is onsite.6.0- Support-Documents 6.1 Form 1, Medical Emergency Information Data 6.2 Data Sheet 1, Acting Security Manager Checklist 6.3 Data Sheet 2, Security Briefing Checklist OERPO1-ZV-.TSO8 Rev. 16 Pa ge 3 of 20*Security Manager 3.2 OERPO1-ZV-IN05, Site Evacuation is required at a Site Area Emergency Classification or greater unless to do so would put site personnel at risk. Site Evacuation may be ordered by the Emergency Director at anytime as dictated by conditions.

3.3 The Technical Support Center is activated at an Alert Emergency or higher classification in accordance with Procedure 0ERP01-ZV-IN01, Emergency Classification.

3.3.1 The Emergency Director has ordered the activation of the Technical Support Center to support response activities.

4.0 References

4.1 STP Emergency Plan 4.2 0ERPO I -ZV-INO 1, Emergency Classification

4.3 OERPO

1 -ZV-IN03, Emergency Response Organization Notification 4.4 -ERP0!-ZV-IN04,.Assembly and Accountability..............

4.5 0ERP01-ZV-IN05, Site Evacuation 4.6 0ERP01-ZV-SH03, Acting Security Manager 4.7 OERP01-ZV-RE01, Recovery Operations

4.8 OERPO1

-ZV-RE02, Documentation 4.9 OPGP05-ZV-0004, Emergency Plan Implementing Procedure Users Guide 4.10 OPOP04-ZO-0007, Aircraft Crash Onsite 4.11 Security Instruction 2203, Owner Controlled AreaCheckpnints 4.12 NRC Regulatory Issue Summary 2009-10, Communications Between the NRC and Reactor Licensees During Emergencies and Significant Events.5.0 Procedure 5.1 At an Alert or higher Emergency Classification or as directed by the Emergency Director report to the affected Unit's Technical Support Center and implement Data Sheet 1, Step 1.0 Initial Activities.

5.2 Complete

Checklist activities as follows: 5.2.1 Use the right column to log the time an activity is performed.

HU2 0 OPOP04-SY-0001 Rev. 8 Page 2 of 30 Seismic Event NOTE.Operational Basis Earthquake (OBE) is defined as "That earthquake which, considering the regional and local geology and seismology and specific characteristics of local subsurface material, could reasonably be expected to affect the plant site during the operating life of the plant; vibratory ground motion for which those features of the nuclear power plant necessary to continued operation without undue risk to the health and safety of the public are designed to remain functional (lOCFRl0OAppendix A)."" Safe Shutdown Earthquake (SSE) is defined as "That earthquake which is based upon an evaluation of the maximum earthquake potential considering the regional and local geology and seismology and specific characteristics of local subsurface material.

It is that earthquake which produces the maximum vibratory ground motion for which certain safety-related structures, systems, and components are designed to remain functional (10CFR100 Appendix A)." (From USFAR Section 2.5.1)* The station design basis values for an OBE is vibratory ground motion equal to or.exceeding a horizontal acceleration of 0.05g, but less than an SSE. The statioh design -basis values for an SSE is vibratory ground motion equal to or exceeding a horizontal acceleration of 0.10g.The accelerometer recorded information can be analyzed and displayed using a personal computer and software supplied with the machine. This software will display the measured response spectrum to be compared with the OBE and SSE response spectrum which will.determine if the OBE or SSE has been exceeded.1.0 Purpose 1'. 1 This procedure provides instructions for determining if a seismic event has occurred, and the appropriate actions to ensure plant safety following an actuation of the Seismic Monitoring System. Instructions are also included for determining if the Operational Basis Earthquake (OBE) or Safe Shutdown Earthquake (SSE) limits have been exceeded.1.2 This procedure is applicable in all modes.2.0 Symptoms and Entry Conditions 2.1 (Unit 1 Only) "SEISMIC EVENT" alarm. (Lampbox 9M01, Window E-8)2.2 (Unit 1 Only) "SEISMIC TRIGGER" triggers the "SEISMIC EVENT" alarm.2.2.1 "SEISMIC TRIGGER" -This alarm indicates that an acceleration signal greater than 0.02g in the vertical or horizontal direction has been detected from the RCB Foundation seismic trigger accelerometer (0-SY-XR-00 11, -37 ft RCB Tendon Gallery AZ 2950).2.3 Physical symptoms of a seismic event have been observed. (e.g., ground motion felt by plant personnel).

NOTE Determination of OBE or SSE should be complete within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.4.4 DIRECT I&C personnel to retrieve all seismic instrumnentation data per OP SP02-SY-0012, Seismic Monitoring Data Retrieval and System Functional Test.4.5 DIRECT I&C to provide ERO TECH Support I&C Engineers with data.4.6 DIRECT ERO TECH Support I&C Engineers to determine if OBE or SSE was exceeded per OPSP09-SY-0001, Seismic Monitoring Data Analysis.NOTE The infornation-from-the National Earthquake Information Center is just for confirmation-that a Seismic Event or strong explosion happened; not to be used to determine OBE or SSE unless on-site instrumentation not available.

4.7 CONTACT

the National Earthquake Information Center in Denver Colorado at phone number (303) 273-8500 (voice) or (303) 273-8516 (tape) for confirmation that a seismic event has taken place.NOTE Modified Mercalli Intensity VI as defined by the USGS: Felt by nearly everyone; many awakened.

Some dishes, windows broken. Unstable objects overturned.

Pendulum clocks m saystop,_* National Earthquake Information Center will list magnitude of earthquakes and distance from nearby cities in km.4.8 IF Seismic Monitoring Data can NOT be obtained in Step 4.6, THEN any one of the following can be used as confirmation of exceeding an OBE: " The earthquake resulted in Modified Mercalli Intensity VI or greater within 5 km (3.1 miles) of the plant,* The earthquake was felt within the plant and was of magnitude 6.0 or greater,* The earthquake was of magnitude 5.0 or greater and occurred within 200 km (124.3 miles) of the plant.

HU3

> ...STPEGS UFSAR 3A4 WATER LEVEL (FLOOD) DESIGN The methods of analyais used to determine the design basis flood are discussed in Section 2.4, These methods are consistent with the requirements.

ofRegulatory Guide (RG) 1.59.The protection measures used to accommodate static arid dynamic flood loads on Category I structures generally fall under the category of "Incorporated barriets" as specified in regulatory positibn C. I of RG' 1.102..3.4.1 Flood Protection 3,4.1.1 &teral Flood Proteotjon Measures for Ses*ic Oatugory I Stmutures, The flooding due to a postulated Main Cooling Reservoir (MCR) embankment breach produces the maximum water level around the power block straotires as well as the oontr6lling water elevations fbr buoyancy calculations.

This. is also the controlling phenomena in deteraiilning the maximum water level at the Esseantial Cooling Water Intake Sttucture

-(WISý. Studies and analyses on the MCR embanledient have demonstrated that an adequate margin of safety'oan be maintained for all credible failure mehmaalams (Section 2,5.6). Accordingly, mechanistic effects (such as scour and.-- eesion-)-asaoakted.-w4.th.

a-postulated-

--.-- --The maximum water level on a vertical face at the south end of the plant structures is El.. 50,8 ft mean sea level N4SLJ5 which is El. 22.8 ft above plant grade., This maximum elevation occurs during a quasi-stoady-#tate condition after a breach of the MCR embanulient and is based on an instantaneous removal of alpproximatoly 2,000 ft of the embankent opposite the power block sructures, This maximum elevation occurs on. the. souih f~ce of the Fuel-Handlhing Building .(FHI) of Unit 1. The selection of postulated embankment breach widths and the, asvmptions mnade in determining the maximum flood elevations, are described in Section 2A.4.4.Total intidation of the Essential' Cooling Pond (ECP) occurs only under the conditioon of MCR embankment breach and 'does not affect the safe shutdown capability of th6 plant, The maximum water level ohlculated'to oocurat the BCWIS is El. 40.8 ft.Safey.4d1ated stniutdifes, q~es af~ind, comxpotifiets Nistd 1&Thble 3Z2A-K'Iare protectd-aginsft_&

-effects of external flooding by: I. Being designed to withstand the maximum flood level and aisooiated effects and temain functional (such as seismic Category I -structures and the Category I auxiliary feedwater-storage-tank):or

' ;2. ' Being housed .within seismic Category I structures which are designed as in item 1, above.Flood protectlon of safety-related structures, systems, and components is provided for postulated

'I flood levels and conditions described fn Section 2.4..4 Qi., Seismic Category I structures are designed to withstand the maxtimum flood le~'els by: ].3.4-1.Revision 13.

,(.]: STFBGS MFSAR 1. Having external walls and :slabs of structures designed to resist the hydrostatic and hydrodynamic foroes associated with surge-wave runup and steady-state water level, 2. Ensuring the overall, stability -of thb total structure against overturning and sliding due to the hydrostatic and hydrodynamic forces associated with surge-wave runup and steady state water level, and 3. Enisuring that the total structure will not float due to buoyancy forces.Figure 3.4-1 shows. a general section through the plant. Figure 3-A-2 shows the seismic. Category I Building xnxlmum steady-state water surface profile, and the corresponding relationship of sill elevations for entrances to seismic Category I buildings.

Table 3A4-1 shows the results of hydraulic loading and buoyancy calculations which were done for the various safety-related facilities, 'The water depths shown on this table were developed from the maximum water surface elevations presented in Table 2AA-3.An investigation of seismic Category I struotures has been made for the flood levels and-associated.

factors greater than 1.1. All exterior seismic Category I building openings are. located above the maximur steady-state flood level or are equipped with watertight doors when located below this" profile, except as' stated below, Exceptions to the above-stated design basis for exterior building, openings in seismic Category I structures are: (1) the opening for the truck bay in the radwaste loading area of the Mechanlcal..

Electrical Auxiliaries Building (MEAB) and (2) the opening for the rail oar 4aoess in the sent f&el.cask loading area of the FHB. These areas are not protected from flooding because they do not have an4 safety-related systems and components located near or below the maximum flood level which is required to perform ahy essential function.

In addition, the two -areas are separated from the remalnd& of the building by wails which do not contain openings below the maximum water surface-elevation corresponding to their location.

The Tendon Gallery Access Shaftoover (TGAS) is-pro~~~ ... .. -. n-The safety-related equipment in the ECWIS is protected from the effects of the design basis flood, The porsonnel access doors on the west wall are piovided with watertight doors; all other doors and openings are above the flood level, The dividing walls and doors between the ECWIS compartmenits minimize the potential for the propagation of flooding from one compartment to another.The three malnt6nance Imookout panels in the. exterior walls of the Diesel-Generator Building .(DOB), which, are located below the maximum water surface elevqtion of 45,0 fL MSL, are watertight and designed for the hydrostatic forces. Bach kmockout panpl allows aoeoss to only one of the three separate compartments within the structure, hnd. only one panel may be removed at one time, The dividing walls between the compartments preclude propagation of flooding from oae compartment to (.. j, 'another.3.4-2 Revision 13 STPEGS UFSAR The maintenanoe knockout panels in the exterior wall.of the room, housing the componont cooling water heqt exchangers in the MEAB are located below the maximum steady-state water level shown I on Figure 3,4-2. These panels are watertight.

Since mechanistic effects fronfi the MCR breach need not be evaluated, there is adequate time to. replace the knockout panels for theý remaining flood events of conceni.All exterior seism'ic Category I building wall and slab surfaces below grade are waterproofed.

This conservatively protects the substructure of seismio Category I buildings from. .groundwater, which is expected.

to stabilize between El. 17 ft and 26 ft (1 to 10 ft below grade) after decommissioiaing of the dewatering system. No waterproofing Is provided on exterior wall or slab surfaces above grade to protect against the effeots, of surge-waver mn-up because of its short duration.

All construotionjoluts in exterior walls and slabs (except for localized areas of blockouts) are provided with waterstops to the maximum flood level for that location -and can withstand hydrostatic and hydrodynamlo effects, All seismic joints betwieen Category I structures contain dual 9-in, water stops bapable .of withstanding potential seismic: and hydrostatic effects. Cracks ia concrete are: mnimized by imposing strict QA and QC prooedures oh the quality of concrete and onstruction techniques.

• 'DfaiffY~-f6 pf6vid~d Wl cka dhliainhWF~kteihiafl6ddi~fo~rThfii~teu -ri1tt~h rn --flooding through the inadvertent introduction of water thrdugh these drains into sesmio Category I structures, The duct býnks are'sealed so as to prevent backflow into safety-related areas, The cable in the duct banks Is designed/specified for submerged installations..

Leakage from groumdwater Into the FHB is prevented by the use of on exterior wall and slab surfaces located below grade. Should groundwater inleakage occur, if fs handled by the ptunps in the FHB stnnp; the three-train compartment siunps, and the transfer cart area sump. For Unit 1 only, accumulated groundwater inieakage to the 64 degree tendon buttress area drains through a penetration in the RCB, tendon gallery outer will and is. collected Itn the tendon gallery sump.-UV o -and slab surfaoes located below grade, Should grgundwater leakage occur, iJ'will be collected in sumps, Discharge from non-radioactive sump,% are routed to.the reservoir via a circulating water discharge line, Potentially radioactive discharge is pumped to the Liquid Waste Processing System (LWPS).3.4,2 Analysts Procedures 3.4.2.1 Phenerena Considered in Deskn Load Calculations, For external flooding, the design basis events considered in design load calculations afe as described in Section 3.4.1.3.4,2'.2 Flood-Force Apxlicatlon.

The design flood conditions and elevations have been... determinedtfrom an analysis of the.phenomena discussed in Section 3.4,1,., 3.4-3 Revision 13 STPEGS PSFEAR In order to establish the controlling load conditions resulting from the embankrment breach, both Instantaneous surge wave runup as well as the longer term, quasi-steady-state -conditions were analyzed.

Tlhe wav6 runup condition conservatively assumes that the maximum total force perpendicular to the south face of the plant structures Includes a dynamic component in addition to the associated hydrostatic fproes, The quasi-steady state condition assumes that only the hydrostatic component cottributes to the.development of the total force for this case, The latter condition resulted im higher water surface elevatiotis and greater hydraulic loads on power block structures.

The vertioal buoyant loading condition is the force equal to the weight of water displaced by a structure.

The discussion of lateral and vertical loadings is presented in the following subsections.

Table 3.4-1 shows a summary of different lateral loadings at various locations aroundplant and ECP structures, caused by their respective controlling flood conditions, Procedures used to determine flood loadings are.Identified in Sections 3.4.2,2.1 and 3A42,2,2.3.4,2.2.1 Lateral Loading: 3.4.22.1,1 , Lateral LoadIh on the Power Block

-Thb analysis of the lateral force on the jower block structures considered both the instantaneous wave runup and the quasi-steady state conditions.

This analysis determined that the maximum total lateral force on the power.-

when the-maximum-water-:level isreaohed'durlng-the-quasi-steady-s'taer, condition; Table 3.4-1 shows the controlling lateral forces (hydrostatic).

exerted on different power block structures.

These lateral forces are treated as triangular loadings on a vertical surface, valring.at a rate of 62.4 lb/t2/ft of structure dep1h, The procedures used to determine the dynamic taid hydrostatic loadings for the above analysis condlitions are discussed below: 1, Dynamic Force The dynamio force on the south side of the power block structures is determined by application of.linear nioomentum principl.s, The flow from the MCOR is assumed to be normal to the south side of the power block structures.

Therefore, the dynamic force exerted on the structures can be exprdssed by the. following momentum equation (Ref, 3.4-2), where: V =dynamic force normal to plant structure P c density of flow_ flo-w-rate V 0 = velocity of flow The maximum value .ofpQv 0 during surge formation is calculated.

This is the contilbution of momentum flux to the dynamic foroo. The. contribution of the unsteadiness -of momentu-n field is insignificant.v

2. H-tydirostatic Force 3,4-4 Revision 1.3