Text

Response to Request for Additional Information Regarding Request for Licensing Action on Alternative Accident Source Term Radiological Dose Calculations
	ML14203A625
Person / Time
Site:	Perry
Issue date:	07/22/2014
From:	Harkness E J; FirstEnergy Nuclear Operating Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	L-14-245, TAC MF3197
	Download: ML14203A625 (93)
	v • d • e

FENOCŽ RrstEnergy Nuclear Operating Company Em8st J. Harkness Vice President July 22, 2014 L--*14-245 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

SUBJECT:

Perry Nuclear Power Plant Docket No. 50-440, License No. NPF-58 Perry Nuclear Power Plant P.O. Box97 10 Center Road Perry. Ohio 44081 440-280-5382 Fax: 440-280-8029 10 CFR 50.90 Response to Request for Additional Information Regarding Request for Licensing Action on Alternative Accident Source Term Radiological Dose Calculations (TAC No. MF3197) A request for licensing action regarding alternative accident source term radiological dose calculations was submitted to the Nuclear Regulatory Commission (NRC) by letter dated December 6, 2013 (Accession No. ML 13343A013).

The NRC staff requested additional information in a letter dated June 24, 2014 (Accession No. ML 14162A409).

The requested information is provided in Attachment

1. In addition, the effects of several necessary changes to the supporting dose calculations are presented in Attachment

2. No changes were identified to the previously provided Significant Hazards Consideration.

There are no regulatory commitments contained in this submittal.

If there are any questions or if additional information is required, please contact Mr. Thomas A. Lentz, Manager-Fleet Licensing, at 330-315-6810.

I declare under penalty of perjury that the foregoing is true and correct. Executed on July 2Z. , 2014. Sincerely, Ernest J. Harkness Attachments:

1. Response to Request for Additional Information on Radiological Dose Calculations

2. Effects of Necessary Changes to Supporting Radiological Dose Calculations cc: NRC Region Ill Administrator NRC Resident Inspector NRC Project Manager State of Ohio (NRC Liaison) Utility Radiological Safety Board "\'.,

Attachment 1 L-14-245 Response to Request for Additional Information on Radiological Dose Calculations Page 1 of 16 FirstEnergy Nuclear Operating Company (FENOC) provided a request for licensing action (RLA) regarding alternative accident source term radiological dose calculations to the Nuclear Regulatory Commission (NRC) in a letter dated December 6, 2013. The NRC staff requested additional information in a letter dated June 24, 2014 .. The requested information is identified using bold text, followed by the FENOC response.

1. For those changes to the current licensing basis (CLB) parameters used in the affected dose consequence analyses, provide additional Information describing for each affected design basis accident, all the basic parameters used in the dose consequence analyses.

For each parameter, please indicate the CLB . value, the revised value where applicable, and the basis for any. changes made to the CLB values. The U.S. Nuclear Regulatory Commission (NRC) staff requests that this information be presented in separate tables for each accident evaluated.

Response:

A matrix is provided for the loss of coolant accident (LOCA), the control rod drop accident (CRDA) scenarios, and the main steam line break outside containment (MSLBOC) analyses.

LOCA Parameter Current Licensing Basis Proposed Licensing Basis for changes [Existing location]

Basis [Proposed location]

Core Source Term General Electric (GE)12/ Global Nuclear GNF2 fuel design to Basis GE14 Fuel (GNF)2 be introduced at [GE12= Amendment 112 [Updated Safety Analysis PNPP during next (power uprate), Report (USAR) fuel cycle NRC Safety Evaluation pg. 15.6-65] (SE) pg. 4; GE14= 10 CFR 50.59 review when converted from GE12] Power Level 102% of rated thermal 102% of rated thermal No change power = 3758 megawatts power = 3833 MWt (existing LOCA thermal (MWt) x 1.02 = [USAR pg. 15.6-24 & dose calculation in 3833 MWt 15.6-58 thru 60] support of [Amendment 112; NRC SE Amendment 112 did pg. 45] assume 3833 MWt although several USAR pages simply reflect the licensed 100% RTP value of 3758 MWt)

Attachment 1 L-14-245 Page 2of16 Isotopes in source term Dose Conversion Factors Gap Release Timing Credit for decay during 2 minutes before start of the gap release. Suppression Pool Scrubbing Drywell Volume Containment Volume (excluding drywell} Volume of Sprayed Region Volume of Containment Unsprayed Region Volume of Total Unsprayed Regions (including drywall} 76 isotopes [Calculations supporting Amendment 1 a3 (the PNPP pilot plant AST amendment}]

Federal Guidance Report (FGR) 11 [USAR pg. 15.a-37) 3a seconds [This assumption was not specified in the USAR but was assumed in the calculations supporting Amendment 1 a3] Not considered No Credit [USAR 15.6-27) 2. 765x1 as feet (ft}3 [USAR pg. 15.6-58] 1.1654x1 as ft 3 [USAR pg. 6.5-56 & 15.6-58) 4.81x1as ft 3 [USAR pg. 15.6-59) 6.84x1as ft 3 [USAR pg. 15.6-59) 9.6a7x1 as ft 3 [not currently specified separately in the USAR] 6a isotopes Based on isotopes [Request for licensing used in the action (RLA} submittal RADTRAD dated 12/6/2a13, and computer code Amendment TBD dated TBD] FGR 11and12 Use of Regulatory

[USAR pg. 15.a-37) Guide (RG} 1.183, Section 4.1.2 and 4.1.4. 2 minutes (BWR) Per RG 1.183, [USAR pg. 15.6-22 is Table4; revised to reference BWR-specific RG 1.183 for timing of the releases, and Table 4 of RG 1.183 specifies 2 minutes for onset of BWR gap release] Decay considered Utility decision since [USAR pg. 15.6-23) RG 1.183 does not specify this No Credit No change [USAR 15.6-27) 2. 765x1 as ft 3 No change [USAR pg. 15.6-58) 1.1654x10 6 ft 3 No change [USAR pg. 6.5-56 & (Note: Table 6.5-9 & 15.6-58) Table 15.6-12a markups are correcting typos) 4.812x1as ft 3 Essentially no [USAR pg. 15.6-59) change (third decimal point} 6.842x1as ft 3 Essentially no [USAR pg. 15.6-59) change (third decimal point} 9.6a7x1 os ft 3 No change [USAR pg. 15.6-59)

Attachment 1 L-14-245 Page 3of16 Flow Rate from Drywall to Unsprayed Region of the Containment Flow Rate from Unsprayed Region of the containment back to Drywell Primary Containment Leakage Rate (sprayed and unsprayed regions leakage) Percent of Primary Containment Leakage that goes Into the annulus Percent of Primary Containment Leakage that bypasses the Secondary Containment Credit for Reduction in Primary Containment Leakage based on containment pressure after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 0-2 hours (hrs.) 3000 cubic 0 -0.5 hour5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months (hr.) 0 cfm feet per minute 0.5-2 hrs. 3000 cfm (cfm) 2 -720 hrs. 2.77x10 5 cfm 2_720 hrs. 2_7 , 7 x 105 cfm [USAR Table 15.6-12b on [USAR Table 15.6-12b on pg. 15*6-59 1 pg. 15.6-59] 0 -2 hrs. 0 cfm 2 -720 hrs. 2. 77x10 5 cfm [USAR pg. 15.6-59] 0.2 percent of containment atmosphere per day (La) [USAR pg. 6.2-72 & 73; Technical Specification (TS) Section 5.5.12] 89.92 percent (0.8992 La) [USAR 15.6-31] 10.08 percent (0.1008 La) [USAR pg. 15.6-31] Not taken 0 -2 hrs. O cfm 2 -720 hrs. 2.77x10 5 cfm [USAR pg. 15.6-59] 0.2 percent of containment atmosphere per day (La) [USAR pg. 6.2-72 & 73; TS Section 5.5.12] 89.92 percent (0.8992 La) [USAR 15.6-31] 10.08 percent (0.1008 La) [USAR pg. 15.6-23 & 31) Based on containment pressure reduction after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , leakage is reduced to 69 percent of the value used during the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months [USAR pg. 15.6-30] No actual change (Note: USAR Table 15.6-12b did not reflect the zero flow rate out of the drywall during the first 0.5 hour5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months period that was in the existing calculation)

No actual change (USAR markup of page 15.6-59 simply rounds 2. 765 to 2.77 x 10 5 cfm, which is the value in the existing calculation)

No change No change (Rev. 0 of new LOCA calculation used 100 percent, but Rev. 1 corrects that to 89.92 percent) No change Credit taken for reduction in containment pressure, as permitted In RG 1.183. See pages 29 & 30 of the RLA Degree of Conformance matrix Attachment 1 L-14-245 Page 4of16 Credit for Reduction In Secondary Containment Bypass Leakage based on containment pressure after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Containment Spray Initiation Containment Spray Duration Spray Fall Height Containment Spray-Aerosol (particulate)

Removal Other Inputs to Spray Modeling Not taken 1 O minute auto initiation (based on high pressure and LOCA signal) [USAR 15.6-32] 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months [USAR pg. 6.5-12 & 15.6-32] 53.2 feet [USAR pg. 15.6-59) Sprayed region of the containment modelled using STARNAUA methodology

[USAR pg. 6.5-13, 14, 56, & 58) (Note: In Am. 103, NRC used RADTRAD goth percentile uncertainty distributions, per pg. 10 of the NRC SE for Am. 103) Various (See USAR Table 6.5-9 "Input Parameters For The Spray Removal Analysis" [USAR pg. 6.5-56] After 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , leakage is Credit was taken for reduced to 69 percent of reduction in the value used during the containment first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months pressure, as (USAR pg. 15.6-30) permitted in RG 1.183 30 minute manual initiation Initiation time is [USAR 15.6-32) conservatively tripled, because. auto-initiation on containment high pressure might not be achieved 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months No change [USAR pg. 6.5-12 & 15.6-32] 54.05 feet New calculation

[USAR pg. 15.6-59) Sprayed region of the Conservative containment modelled RADTRAD model using RADTRAD Powers assumption, model with 10 1 h percentile RG 1.183, uncertainty distribution.

Appendix A, Particulate removal Section 3.3 coefficient due to sprays is reduced by a factor of 10 after the aerosol mass is depleted by a factor of 50 [USAR pg. 6.5-13, 14, & 56) Q = 0.0621 (Spray Flux) Powers spray model Alpha = 1.422 (Ratio of inputs unsprayed volume to sprayed volume) [USAR pg. 6.5-56)

Attachment 1 L-14-245 Page 5of16 Containment Spray-Elemental Iodine Removal Natural Deposition-Aerosol (particulate)

Removal Natural Deposition-Elemental Iodine Removal Engineered Safety Feature (ESF) Leakage Pathway Direct to Environment Offsite Breathing Rates (meters (m)3/ second (sec)) Sprayed region of the containment modelled using STARNAUA methodology

[USAR pg. 6.5-13, 14, 56, & 58) Drywall region modelled using STARNAUA methodology

[USAR pg. 15.6-28 & 29) (Note: In Am. 103, NRC used RADTRAD 90lh percentile uncertainty distributions, per pg. 7 of the NRC SE for Am. 103) Drywall region modelled using STARNAUA methodology (USAR pg. 15.6-30) Time (hours) Leak Rate 0-24 15 gallons per hour (gph) 24-24.5 15 gph + 50gpm 24.5-720 15gph [USAR pg. 15.6-33 & 60] Time (hours) 0-8 3.47x10 4 8-24 1.75x10 4 24-720 2.32x10 4 [USAR pg. 15.0-37) Sprayed region of the RG 1.183 and containment modelled SRP6.5.2 using Standard Review Plan (SRP) 6.5.2 guidance.

Elemental Iodine removal by sprays is terminated when a OF of up to 200 is reached [USAR pg. 6.5-13, 14, & 56) Drywall and unsprayed Conservative volume of containment RADTRAD "'odel modelled using RADTRAD assumption Powers Model for aerosol removal with 10 1 h percentile uncertainty distribution

[USAR 15.6-28 & 29) Drywall, sprayed, and RG 1.183 and unsprayed volume of SRP6.5.2 containment modelled using SRP 6.5.2 guidance.

Similar to the spray assumptions, elemental Iodine removal is terminated when a OF of up to 200 is reached [USAR pg. 6.5-13, & Table 6.5-11) Time (hours) Leak Rate No change 0 .. 24 15gph 24-24.5 15 gph + 50gpm 24.5-720 15gph [USAR pg. 15.6-23, 33 & 60] Time (hours) RG 1.183, 0-8 3.5x10 4 Section 4.1.3 8""'.'24 1.8x10 4 24-720 2.3x10 4 [USAR pg. 15.0-37)

Attachment 1 L-14-245 Page 6of16 Main Steam Line (MSL) leak rate (standard conditions)

! Main Steam Line leak rate converted from standard to non-standard conditions (assuming elevated post-accident temperatures) during first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Credit for Main Steam line leakage reduction associated with decreased pressure after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . 250 standard cubic feet per hour (scfh) total, 100 scfh maximum per line (under standard conditions).

Modelled as 100 scfh to the broken MSL, and 150 scfh to the intact MSL's [TS Surveillance Requirement (SR) 3.6.1.3.1 O] Flow rate from drywall to all steam lines (both broken and intact) = 298 cfh for first 7 484 seconds, and 247 cfh until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Flow rate between MSIVs in one steam line = 191 cfh (Note: USAR Table 15.6-12a says cfm vs. cfh, which was a typographical error) [USAR pg. 15.6-58] Not taken [USAR 15.6-25 & 58] 250 scfh total, 100 scfh No change maximum per line (under standard conditions).

Modelled as 100 scfh to the broken MSL, and 150 scfh to the intact MSL's [TS SR 3.6.1.3.1 O] Flow rate from drywall to No change the broken steam line (values in USAR = 1. 987 cfm for first 7 484 Table 15.6-12a seconds, and 1.647 cfm appear to be until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months changed, but that is Flow rate from drywell to due to mathematical the intact steam lines conversion of . = 2.98 cfm for first 7484 existing assumed seconds, and 2.47 cfm flow rates from cfh until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to cfm (engineering Flow rate in one steam unit change only), line, between the MSIVs and inclusion of = 3.183 cfm more detail into this Flow rate in the other (the table about how intact) steam lines, flow rates to (and between the MSIVs through) the main =4.775 cfm steam lines are [USAR pg. 15.6-58] attributed to the broken versus the intact main steam lines) After 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , leakage Some credit was from the drywell to the taken for reduction Main Steam Lines is in containment reduced to 69 percent of pressure, as the value used during the permitted in first 24 hrs. However, the RG 1.183 leak rate assumed between the MSIVs is conservatively not reduced after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . [USAR pg. 15.6-25 & 58]

Attachment 1 L-14-245 Page 7of16 Aerosol Removal Efficiencies in Steam Lines Annulus Exhaust Gas Treatment (AEGT) system recirculation to the annulus AEGT Filtration (High Efficiency Particulate Air (HEPA) filter) Broken steam line -between MSIVs (100 scfh) -For Particulate Iodine, solubles, and insoluble 68.1 % from 0 to 0.5 hr. 83.5% from 0.5 to 1.5 hrs. 87.13% from 1.5 to 3 hrs. 89% from 3 to 5 hrs. 86.14% from 5 to 7 hrs. 81.85% from 7 to 9 hrs. 76.9% from 9 to 11 hrs. 36.53% from 11 to 720 hrs. -For Elemental Iodine -45% Intact Steam lines RPV to IB MSIV (150 scfh) -For Particulate Iodine, solubles and insoluble 72.06% from 0 -end -For Elemental Iodine -45% Intact Steam Lines between MSIVs -For Particulate Iodine, solubles, and insoluble 71.4% from Oto 0.5 hr. 81.3% from 0.5 to 1.5 hrs. 83.61 % from 1.5 to 3 hrs. 84.49% from 3 to 5 hrs. 83.39% from 5 to 7 hrs. 79.98% from 7 to 9 hrs. 75.58% from 9 to 11 hrs. 38.07% from 11 to 720 hrs. -For Elemental Iodine -45% [USAR pg. 15.6-26 & 27 provide general discussion of steam line treatment]

Not considered, such that flow rate to the environment

= 2000 cfm [Not currently specified in the USAR] 99 percent efficiency

[USAR pg. 15.6-31 & 59] Broken steam line -No change between MSIVs (100 scfh) . -For Particulate Iodine, solubles, and insoluble 68.1 % from 0 to 0.5 hr. 83.5% from 0.5 to 1.5 hrs. 87.13% from 1.5 to 3 hrs. 89% from 3 to 5 hrs. 86.14% from 5 to 7 hrs. 81.85% from 7 to 9 hrs. 76.9% from 9 to 11 hrs. 36.53% from 11 to 720 hrs. -For Elemental Iodine -45% Intact Steam lines RPV to 18 MSIV (150 scfh) -For Particulate Iodine, solubles and insoluble 72.06% from 0 -end -For Elemental Iodine -45% Intact Steam Lines between MSIVs -For Particulate Iodine, solubles, and insoluble 71.4% from 0 to 0.5 hr. 81.3% from 0.5 to 1.5 hrs. 83.61 % from 1.5 to 3 hrs. 84.49% from 3 to 5 hrs. 83.39% from 5 to 7 hrs. 79.98% from 7 to 9 hrs. 75.58% from 9 to 11 hrs. 38.07% from 11 to 720 hrs. -For Elemental Iodine -45% [USAR pg. 15.6-26 & 27 provide general discussion]

Not considered, such that No change flow rate to the environment

= 2000 cfm [USAR pg. 15.6-31] 99 percent efficiency No change [USAR pg. 6.5-4; 15.6-31 & 59) ,-.. : .*,'"

Attachment 1 L-14-245 Page 8of16 AEGT Filtration (Charcoal)

Exclusion Area Boundary (EAB) X/Q ( sec/m 3) Low Population Zone (LPZ) X/Q (sec/m 3) Control Room X/Q (sec/m 3) Control Room Volume Control Room Filtration System (HEPA) Control Room Filtration System (Charcoal)

Zero percent efficiency

[USAR pg. 6.5-3) 4.3x10-4 [NRC Safety Evaluation for Amendment 103, pg. 15) Time (hrs) 0-8 4.8x1Q*S 8-24 3.3x1Q*S 24-96 1.4x1Q*S96-720 4.1x10.a [NRC Safety Evaluation for Amendment 103, pg. 15) Time (hrs) 0-8 3.5x10-4 8-24 2.1x10-4 24-96 1.1x10-4 96-720 5.75x10-5 [NRC Safety Evaluation for Amendment 103, pg. 14) 3.44x10 5 ft 3 [USAR pg. 15.6-64) 95 percent credit for HEPA removal of Particulates.

[USAR pg. 15.6-35) 50 percent charcoal filter efficiency for elemental and organic iodine. [USAR pg. 6.5-4, 15.6-23 & 35) Zero percent efficiency No change [USAR pg. 6.5-3 & 42) 4.3x10-4 No change [USAR pg. 15.6-63) Time (hrs) No change 0-8 4.8x10-5 8-24 3.3x10-5 24-96 1.4x10-5 96-720 4.1x10.a [USAR pg. 15.6-63) Time (hrs) No change 0-8 3.5x10-4 (markups on USAR 8-24 2.1x10-4 page 15.6-63 are 24-96 1.1x10-4 correcting incorrect 96-720 5.75x10-5 information)

[USAR pg. 15.6-63 with corrections to reflect correct CLB] 3.90x10 5 ft 3 Calculation for [USAR pg. 15.6-64) control room volume revised since pilot plant submittal 99 percent credit for HEPA Acceptable value removal of Particulates based on current [USAR pg. 6.5-4; 15.6-35) and future TS test acceptance criterion 80 percent charcoal filter Eliminated an efficiency for elemental unnecessary and organic iodine. conservatism;

[USAR pg. 6.5-4 & 30, 80 percent value will 15.6-23 & 35) support a future request for licensing action to revise TS charcoal adsorber testing acceptance criterion to a value of 10% penetration

(=90% efficiency) versus the current 2.5% penetration requirement Attachment 1 L-14-245-Page 9of16 Control Room Emergency Recirculation System ActuationfTiming of the Start of Filtration Timing of the Reduction of Unfiltered lnleakage Into the Control .Room Control Room Breathing Rate Control Room Occupancy Factors 30 minutes [USAR pg. 6.4-8, 15.6-23) 1375 cfm unfiltered inleakage starting at time (t)=O, for duration [USAR pg. 6.4-14 & 15, 15.6-35 & 64) 3.47x10-4 m 3/sec [USAR pg. 15.0-37) Time (hours) 0-24 1.0 24-96 0.6 96-720 0.4 [not currently specified in the USAR] 30 minute manual operator No credit taken for actuation automatic actuation of filtration or for [USAR pg. 6.4-8, 14, & 15, any reduction in 15.6-23 & 35 unfiltered In-leakage until 30 minutes, to support a future request for licensing action to remove TS controls over auto-initiation instrumentation 6600 cfm unfiltered Conservatively inleakage from t=O to assumes outside air 30 minutes inlet is not isolated 1375 cfm from t=30 min to until t=30 minutes when Control Room 30 days Emergency

[USAR page 6.4-14 & 15, Recirculation 15.6-35 & 64) System is manually actuated, to support a future request for licensing action to remove TS controls over auto-initiation instrumentation 3.5x10-4 m 3/sec RG 1.183, [USAR pg. 15.0-37) Section 4.2.6 Time (hours) No change 0-24 1.0 24-96 0.6 96-720 0.4 [USAR pg. 15.6-64) CRDA (Scenario 1 -Standard Review Plan Section 15.4.9) Parameter Current Licensing Basis Proposed Licensing Basis Basis For , [Current Location]

[Proposed Location]

Changes Core Source Term GE14 GNF2 Change in basis [USAR pg. 15.4-27, 45, & [USAR pg. 15.4-27, 45, & 46) fuel design 46]

Attachment 1 L-14-245 Page 10of16 Power Level Coincident loss of offsite power (LOOP)? Isotopes Considered Assumed number of failed fuel rods Iodine Fractions, % Radial Peaking Factor Gap and Melt fraction Activity released to condenser Activity available for release from condenser Plant startup, at low power level, after operation at 3,833MW1 [USAR pg. 15.4-26 thru 29, &44) Yes [USAR pg. 15.4-27 & 28] Considered Iodine, Krypton, and Xenon isotopes [USAR pg. 15.4-29] 1107 [USAR pg. 15.4-27, 28, & 44] Organic 0 Elemental 100 Particulate 0 [USAR pg. 15.4-44] 1.7 [Reload Analysis Parameter]

Group Gap Melt Noble Gas 10% 100% Halogen 10% 50% [USAR pg. 15.4-29] Group Noble Gas 100% Halogen 10% [USAR pg. 15.4-29] Group Noble Gas 100% Halogen 10% [USAR pg. 15.4-29] Plant startup, at low power level, No change after operation at 3,833 MW1 [USAR pg. 15.4-26 thru 29, & 44) Yes No change [USAR pg. 15.4-27 & 28] (per SRP 15.4.9 Section Ill, Assumption

1) Considered Iodine, Krypton, Added Xenon, Bromine, Cesium, and Isotopes per Rubidium isotopes RG 1.183, [USAR pg. 15.4-29] Section 3, Tables 3 & 5 1376 (two* full rows of bundles Conservative around the dropped control rod) increase In number of [USAR pg. 15.4-27, 28, & 44] failed rods Organic 0.15 RG 1.183, Elemental 4.85 Appendix C, Particulate 95.0 Item 3.6 [USAR pg. 15.4-44] 2.0 Current radial [Reload Analysis Parameter]

peaking factor in analyses Group Gap Melt Added alkali Noble Gas 10% 100% metals per Halogen 10% 50% RG 1.183, Alkali Metals 12% 25% Section 3.2, [USAR pg. 15.4-29] Tables 1 & 3 Group RG 1.183, Noble Gas 100% Appendix C, Halogen 10% Item 3.3 Remaining 1% [USAR pg. 15.4-29] Group RG 1.183, Noble Gas 100% Appendix C, Halogen 10% Item 3.4 Particulate 1% [USAR pg. 15.4-29]

Attachment 1 L-14-245 Page 11of16 Condenser Leakage Release to Environment Atmospheric Dispersion X/Q The leakage rate from the condenser is assumed to be 1 % per day for the duration of the event (24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ). [USAR pg. 15.4-29 (24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months termination time is from SRP 15.4.9 Assumption 111.12)] All activity leaking from the condenser is assumed to leak directly to the environment without mixing in the turbine building.

[USAR pg. 15.4-28 & 44] Time EAB LPZ 0-2 hrs. 4.3E-4 4.SE-5 2-8 hrs. --4.SE-5 8-24 hrs. --3.3E-5 1-4 Days --1.4E-5 4-30 Days --4.1 E-6 [USAR pg. 15.4-45] Control Room Dose Not calculated Control Room Emergency Recirculation System Filtration (Charcoal and HEPA) AEGT System Filtration (Charcoal and HEPA) [Control room doses were not previously required to be determined for CRDA] NIA [Control room doses were not previously required to be determined for CRDA] Not

[Releases are from outside containment, so AEGT system is N/A, therefore it is not credited]

The leakage rate from the condenser is assumed to be 1 % per day for the duration of the event (24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ) [USAR pg. 15.4-29] All activity leaking from the condenser is assumed to leak directly to the environment without mixing in the turbine building.

[USAR pg. 15.4-28 & 44] Time EAB LPZ CR 0-2 hrs. 4.3E-4 4.8E-5 3.5E-4 2-8 hrs. --4.8E-5 3.5E-4 8-24 hrs. --3.3E-5 2.1 E-4 1-4 Days --1.4E-5 1.1 E-4 4-30 Days --4.1 E-6 5. 75E-5 [USAR pg. 15.4-45] Calculated

[USAR pg. 15.4-47] Not credited [USAR pg. 6.5-4 & 30, 15.4-45] Not credited [USAR pg. 6.5-3, 4, & 42] No change No change No change except for the addition of control room (CR) X/Q's RG 1.183, Section 4.4 No credit taken No change Attachment 1 L-14-245 Page 12of16 CRDA (Scenario 2 -NED0-31400 11 Safety Evaluation for Eliminating the Bolllng Water Reactor Main Steam Line Isolation Valve Closure Function and Scram Function of the Main Steam Line Radiation Monitor")

Parameter Current Licensing Basis Proposed Licensing Basis Basis For {See Scenario 1, [Current Location]

[Proposed Location]

Changes with the following changes based on NED0-31400}

Power Level Plant startup, at low power Plant startup, at low power level, No change level, after operation at after operation at 3,833 MW1, 3,833 MW1, aligned to the aligned to the Offgas system Offgas system [USAR pg. 15.4-26, 27, 28, & 30) [USAR pg. 15.4-26, 27, 28, & 30) Coincident loss of No, so Offgas System No, so Offgas System remains No change offsite power remains running to hold up running to hold up all but the (LOOP)? all but the noble gases noble gases [USAR 15.4-30) [USAR 15.4-30) MSLBOC Current Licensing Basis Proposed Licensing Basis Basis for Parameter

[Current Location]

[Proposed Location]

Changes Reactor Coolant Original GE-supplied Revised GE-supplied reactor RG 1.183, Inventory reactor coolant inventory coolant inventory for: Appendix D, for: a) Design basis analysis = TS pre-Assumption 2 a) Design basis analysis = accident spike of 4.0 µCi/gm, and TS pre-accident spike of b) Realistic analysis = 4.0 microcuries

(µCi)/ TS maximum equilibrium value of gram (gm), and 0.2 µCi/gm b) Realistic analysis = [USAR pg. 15.6-13 & 15) TS maximum equilibrium value of 0.2 µCi/gm [USAR pg. 15.6-13 & 15] Source Term Considered Iodine, Krypton, Considered Iodine, Krypton, Revised and Xenon isotopes as Xenon, and Bromine isotopes as GE-supplied coolant activity coolant activity reactor [USAR pg. 15.6-52 & 56) [USAR pg. 15.6-52 & 56) coolant isotopes Iodine Fractions Organic 0 Organic 0.15 RG 1.183, (%) Elemental 100 Elemental 4.85 Appendix D, Particulate 0 Particulate 95.0 Assumption

[USAR pg. 15.6-54] [USAR pg. 15.6-54) 4.4 Attachment 1 L-14-245 Page 13of16 Isolation Valve Closure Time, and mass of coolant released to the environment Atmospheric Dispersion X/Q Control Room Dose Control Room Emergency Recirculation System Filtration (Charcoal

& HEPA) AEGT System Filtration (Charcoal and HEPA) Various closure times are listed on current USAR pages, but the resultant mass release (consisting of 14,311 pounds.of steam and 127 ,376 pounds of liquid) listed in the USAR is consistent with the supporting calculation

[USAR 15.6-12 & 13] EAB LPZ 6. 7E-4 8.2E-5 [USAR pg. 15.6-55] Not calculated

[Control room doses were not previously required to be determined for MSLBOC] N/A [Control room doses were not previously required to be determined for MSLBOC] Not credited [Releases are from outside containment, so AEGT system is NIA, therefore it is not credited]

The supporting calculation's.

conservative valve closure time of 6.05 seconds will be reflected in the USAR, with its resultant mass release of 14,311 pounds of steam and 127,376 pounds of liquid [USAR 15.6-12 & 13] Time EAB LPZ CR 0-2 hrs. 4.3E-4 4.8E-5 3.5E-4 2-8 hrs. --4.8E-5 3.5E-4 8-24 hrs. -3.3E-5 2.1 E-4 1-4 Days -1.4E-5 1.1E-4 4-30 Days -4.1 E-6 5. 75E-5 [USAR pg. 2.3-89 & 15.6-55] Calculated

[USAR pg. 15.6-53, 55 & 57) Not credited [USAR pg. 6.5-4 & 30) Not credited [USAR pg. 6.5-3, 4, & 42) Clarification of actual (conservative) valve closure timing in the existing calculation; no change in the amount of reactor coolant released Utilized updated X/Q values and added control room RG 1.183, Section 4.4 No credit taken No change Attachment 1 L-14-245 Page 14of16 2. On page 7 of the submittal evaluation, FirstEnergy Nuclear Operating Company addressed the impact of the proposed*

changes on the radiological habitability of the Technical Support Center. However, the application appears to be silent on the impact of those changes on the calculations supporting the establishment of numeric thresholds for emergency action levels (EALs) related to readings on: (1) radiological effluent radiation monitors, and (2) the in-containment high range radiation monitor readings used in the fission product barrier matrix EALs. In addition to the source term change and core inventory changes, the NRC staff notes that you have proposed changes in other analysis assumptions (e.g., filter removal efficiencies, containment leakage rates, drywell flows, containment spray credit, etc.) that could potentially affect the validity of the previous EAL thresholds.

Explain the impact of the changes on these EAL thresholds.

Response:

The current PNPP Emergency Plan utilizes the Nuclear Utility Management and Resources Council (NUMARC) EAL methodology, NUMARC/NESP-007, Methodology for Development of Emergency Action Levels. When the plant was originally licensed in the mid-1980's, the dose assessment program utilized a source term based on the licensing-basis loss of coolant accident (LOCA) in the Final Safety Analysis Report (FSAR). Subsequently, in October of 1988, the NRC published NUREG-1228, Source Term Estimation During Incident Response to Severe Nuclear Power Plant Accidents.

This guidance reflected a more current understanding of accident source terms and their offsite consequences.

The NUREG noted on page 1-6 that licensing basis evaluations

."normally should not be used to estimate actual accident source terms or offsite doses." The NUREG (beginning on page 2-5) provided core inventory

assumptions to be used for source term estimation, expressed in curies per electrical megawatt (Ci/MWe), such that the source term could be adjusted for various plants. The NUREG explained on page 1-2 that "studies of the uncertainties associated with source term estimation indicate that source term projections based on accident conditions are only accurate within a factor of 100 or more ... " Therefore, the generic, scalable source term provided by NUREG-1228 is considered to provide an acceptable standard for use at various plants, including PNPP. In the early 1990s at PNPP, when dose assessment software was updated, the source term provided in NUREG-1228, scaled for PNPP's licensed power level at that time, was utilized.

The dose assessment program was updated again in 2012, utilizing the Meteorological Information and Dose Assessment Software (MIDAS) program. As noted in NUREG-1228 on page 1-3, "if a change in assumptions does not result in a change to the source term by at least 1 order of magnitude, it is not worth considering

... " Therefore, it was determined that the NUREG-1228 source term, scaled for PNPP's five percent uprate, would be used as input for the dose assessment program (for events postulated to occur with the plant at power). The.dose assessment program was utilized to calculate plant effluent monitor values that would result in 1 Rem Total Effective Dose Equivalent (TEDE) and/or 5 Rem Child Thyroid doses offsite (the current EAL thresholds).

Similarly, for item (2) in the NRC RAI (the in-containment high-range radiation monitors), the PNPP EALs are based on the NUMARC/NESP-007 guidance.

That guidance Attachment 1 L-14-245 Page 15of16 discusses the determination of in-containment radiation monitor readings for fuel failure events, specifically, for the fuel clad barrier on page 5-20 and the primary containment barrier on page 5-23.

5-23 includes a reference to NUREG-1228.

Similar to the discussions above for the effluent radiation monitors, the NUREG-1228 conclusion that unless a change in assumptions results "in a change to the source term by at least .1 order of magnitude, it is not worth considering" can also be applied to the In-containment radiation monitor reading thresholds.

Since the licensing basis source term is not changing significantly with the adoption of the GNF2 fuel, and a NUREG-1228-based source term based on curies per electrical megawatt would not change, it is concluded that the PNPP in-containment radiation monitor reading values specified in the EALs do not need to be changed as a result of the proposed change to the licensing basis source term for the LOCA event. In summary, it is concluded that implementation of a new licensing basis source term, such as proposed by this request for licensing action (RLA), does not invalidate the current numeric thresholds in the Emergency Plan EALs, including the (1) radiological effluent radiation monitors, and (2) the in-containment high range radiation monitors.

Other proposed changes to licensing basis analysis assumptions, such as those listed in the NRC RAI (filter efficiencies, containment leakage rates, drywell flows, containment spray credit, etc.), also would not invalidate use of the current numeric thresholds in the EALs. 3. Address whether any non-safety-related systems and components are credited in the accident source term (AST) analyses.

If so: a. Describe the independence (electrical and physical separation) of these systems from the safety-related systems. Provide a detailed discussion on why a fault on the non-Class 1 E electrical circuit will not propagate to the Class 1 E electrical circuit. b. Describe the redundancy of these systems and how these systems meet the single failure criterion.

Response:

No non-safety-related systems or components are credited to reduce doses in the alternative AST analyses.

Conservative assumptions in several event dose analyses assume the normal control room ventilation system continues to operate for various lengths of time post-accident, which increases the amount of unfiltered inleakage into the control room versus use of an assumption that the systems shut down or otherwise isolate quickly, so this is not considered to be a 'credit.'

The Scenario 1 CRDA analysis notes that if the mechanical vacuum pumps are running at a very low plant power level, the pumps will shut down as a result of the Standard Review Plan 15.4.9-required Assumption 1 of a loss of offsite power (LOOP); this has been previously reviewed and approved by the NRC for the CRDA at PNPP.

Attachment 1 L-14-245 Page 16of16 4. Address whether any loads are being added to the emergency diesel generators (EDGs). If so, describe their Impact on the capability and capacity of the EDGs. Also, describe changes, if any, being made to the EOG loading sequence to support this license amendment request. Response:

No new loads are being added to the EDGs, and there are no changes being made to the EDG loading sequence in support of this amendment request. 5. Discuss why there are no changes to the Environmental Qualification profiles as a result of the full implementation of the AST. Response:

A calculation was performed to determine the impact on equipment environmental qualification due to the post-accident fuel failure radiation dose for GNF2 fuel. The maximum radiation dose rate and the integrated dose (over the course of an . accident) following failure of GNF2 fuel was compared to the maximum dose rate and the integrated dose for the current design basis source term. It was determined that the accident doses from the GNF2 source term trended closely with the original source term, with the GNF2 fuel post-accident doses being slightly higher. The increases for the maximum dose rates were on the order of 10.5 percent, and the integrated dose increases ranged from zero percent to approximately eight percent. The increases represent the expected change in the post-accident dose profiles as a result of the transition to the GNF2 fuel. Equipment Qualification packages were reviewed to determine if these increases would exceed the existing radiation qualification of equipment.

In all cases, the existing qualification doses bound the expected increased doses resulting from the transition to GNF2 fuel. Formal updates to the environmental

qualification zone profiles, equipment calculations, and affected auditable file packages, are ongoing to reflect the results of the above review of the GNF2 post-accident doses.

Attachment 2 L-14-245 Effects of Necessary Changes to Supporting Radiological Dose Calculations Page 1of76 Several necessary changes were identified in two of the calculation summaries included as addendums to the request for licensing action (RLA) submitted on December 6, 2013; specifically Addendum 4, "Summary of Loss of Coolant Accident (LOCA) Dose Calculation," and Addendum 5, "Summary of Control Rod Drop Accident (CRDA) Dose Calculation." Updated summaries of the calculations are included in this attachment, following this discussion (42 pages in the LOCA summary, and 2 pages in the CRDA summary).

The issue of these calculations needing to be changed after their submittal has been entered into the FENOC Corrective Action Program. As a result of the calculation changes, and a re-examination of the proposed Updated Safety Analysis Report (USAR) markups that were included as Addendum 1 of the RLA, updates of the proposed USAR markups are also provided at the end of this attachment, immediately following the CRDA dose calculation update (31 revised or additional USAR pages (the new pages do not include revised USAR text -they are provided only for the purpose of providing information/context)).

The changes to the LOCA calculation include: 1) Correction of the total elemental iodine removal coefficient in the sprayed region of containment to 20.975 hour0.0113 days 0.271 hours 0.00161 weeks 3.709875e-4 months 1 , 2) Correction of the equilibrium source term values for Y90, Y91, Y92, Zr95, Zr97, Nb95, and Mo99, 3) Correction of engineered safety feature (ESF) leakage timing so that ESF leakage is assumed throughout the 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of the analysis, 4) Removal of the unnecessary hydrogen mixing system model (sensitivity), 5) Increase of the unfiltered flow rate into the control room during the first 30 minutes to be consistent with the value assumed in the CRDA and main steam line break analyses (10 percent above nominal system intake flow rate), 6) Corrected containment leakage flow rates into the annulus (secondary containment), and 7) Additional changes to improve readability and flow of information.

The change to the CRDA calculation clarifies the description of the iodine releases assumed in the calculation.

COVERSHEET:

Table of Contents OBJECTIVE OR PURPOSE SCOPE OF CALCULATION TABLE OF CONTENTS SUBJECT

SUMMARY

OF RESULTS/CONCLUSIONS LIMITATIONS OR RESTRICTIONS ON CALCULATION APPLICABILITY IMPACT ON OUTPUT DOCUMENTS DOCUMENT INDEX CALCULATION COMPUTATION (BODY OF CALCULATION):

Page i 1.0 PURPOSE ............

.-....................................................................................................*..................................................

1

2.0 BACKGROUND

...***..*...**.*.*.....*..............*.*..*.*....***.....**..*...*............................*........*.*..**..*..**.**.*..*.**..*...**.....**...*.....***

1 3.0 ACCEPTANCE CRITERIA ***.*...****.*....******.**.****..**.**********..**...*********.***..*****....*.***..******.***********.*********************.******.**********

z 4.0 METHOD OF ANALYSIS *.....*.............*.......*.*...**..*.*...**...**.*........*.*.....*...*..*...*....*..*.********.***********.*****..*......*..*....*****...**

3 5.0 ASSUMPTIONS

...*****.....**.....***............**........*....***....***.......**.....**........*.......*..*...****...****...*.******...***......**...............*...*

3 5.1 CONTROL ROOM EMERGENCY RECIRCULATION SYSTEM (CRERS) *****************************************************************************************************

3 5.2 HYDROGEN MIXING SYSTEM.*************************************************.*******************************************************************************

3 5.3 CONTROL ROOM INLEAKAGE

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

4 5.4 DRYWELLFLOWS

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

4 5.5 CONTAINMENT LEAKAGE RATE ...................................................................................................................................................

5 5.6 AEGTS FILTRATION

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

5 5.7 CONTAINMENT SPRAY ................................................................................................................................... ..........................

5 5.8 ECCS LEAKAGE .......................................................................................................................................................................

6 5.9 MSIV LEAKAGE RATE ***************************************************************************************************************************************************************

6 5.10 BYPASS LEAKAGE .....................................................................................................................................................................

7 5.11 SOURCE TERM RELEASE ************************************************************************************************************************************************************

1 6.0 DESIGN INPUT .......................................................................................

................................................................ **** 8 6.1 PLANT GRADE ..................................................................

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

8 6.2 CORE SOURCE TERMS AND RELEASES ...........................................................................................................................................

8 6.2.J Onset of Gap Release .............................................................................................................................................

JO 6.2.2 Release fractions

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

JO 6.3 SUPPRESSION POOL IODINE RE-EVOLUTION

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

10 6.4 DOSE CONVERSION FACTORS ....................................................................................................... *********************************************

10 6.5 ATMOSPHERIC DISPERSION FACTORS ********************************************** , ********************************************************************************************

11 6.6 BREATHING RATE AND OCCUPANCY FACTORS ******************************************************************************************************************************

. 11 6.7 CONTAINMENTVOlUMES

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

12 6.8 TECHNICAL SUPPORT CENTER DOSES *******************************************************************************************************************************************

12 6.9 MIXING BElWEEN THE UNSPRAYED CONTAINMENT AND SPRAYED CONTAINMENT

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

13 6.10 CONTAINMENT LEAKRATE ..........................................................................................................................................................

13 6.11 LEAKAGE AFTER 24*HouRS ........................................................................................................................................................

14 6.12 MSIV FLOWS *************************************************************************************************************************************************************************

15 6.13 RADIONUCLIDE REMOVAL MECHANISMS

16 Page ii 6.13.1 Removal by Deposition

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

16 6.13.2 Removal by Sprays .............................................................................................................................................

19 6.14 ANNULUS MODEL ....................................................................................................................................................................

24 6.15 DEPOSITION IN MAIN STEAM LINES .............................................................................................................................................

24 7.0 ACCIDENT SCENARIO AND CHRONOLOGY

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

-**--*********oooooo******H******25 8.0* MODEL DEVELOPMENT

- - - - -*---**************************26 8.1 ECCS LIQUID LEAKAGE .............................................................................................................................................................

26 8.1.1 *Source Terms ........................................................................................................................ .................................

26 8.1.2 Volumes ..................................................................................................................................................................

27 8.1.3 Flows ......................................................................................................................................................................

27 8.1.4 Removal Mechanisms

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

27 8.1.S Model ..........................................................................

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

27 8.1.6 Results ....................

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

27 8.2 MSIV LEAKAGE .......................................................................................................................................................................

27 8.2.1 Source Terms ........................................................................................

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

27 8.2.2 Volumes ..................................................................................................................................................................

28 8.2.3 Flows_ ......................................................................................................................................................................

28 8.2.4 Release Points ........................................................................................................................................................

28 8.2.5 Model .....................................................................................................................................................................

29 8.3 CONTAINMENT

& CONTAINMENT BYPASS LEAKAGE ........................................................................................................................

29 8.3.1 Volumes ..................................................................................................................................................................

29 8.3.2 Flows ......................................................................................................................................................................

30 8.3.3 Removal Mechanisms

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

31 8.3.4 Release Points ........................................................................................................................................................

32 8.3.S Model ... : .................................................................................................................................................................

32 8.4 CONTROL ROOM .....................................................................................................................................................................

32 8.5 RADTRAD MODEL ................................................................................................................................................................

33 9.0 OPERATOR ACTIONS ****************************************************************************************************************-******************************33 10.0

- - - - -*****************************36 11.0 OVERALL RESULTS ....................................................................................................................

--*****************************36 ATTACHMENTS Attachment 1: PNPP ESF.out Attachment 2: PNPP LOCA.out Attachment 3: PNPP LOCA TSC.out Attachment 4: PNPP ESF TSC.out SUPPORTING DOCUMENTS (For Records Coav Onlv\ 25 Pages 338 Pages 42 Pages 26 Pages OBJECTIVE OR PURPOSE: Page iii This calculation replaces the current loss-of-coolant accident (LOCA) dose calculation (PSAT 08401T.03, DIN 25). and supports the transition to GNF2 fuel. In addition, certain excess conservatisms contained in the current LOCA dose calculation will be removed to increase the margin of safety. This calculation will be performed in accordance with the guidance provided in Regulatory Guide 1.183 (DIN 7) for application of an alternative radiological source term and will demonstrate that the offsite and onsite accident doses comply with the requirements and acceptance criteria *of 10 CFR Part 50.67. SCOPE OF CALCULATION/REVISION Revision*o PNPP will be transitioning to GNF2 fuel in future outages. The purpose of this calculation is to prepare a dose analysis supporting this transition and to establish the new design basis LOCA dose analysis using the RADTRAD 3.03 computer program, which was developed for the Nuclear Regulatory Commission (NRC) and is in common use for this type of application in the nuclear industry.

This calculation may also be used to support a license amendment request (LAR) associated with the GNF2 fuel transition.

Revision 1 The changes incorporated In Revision 1 are: 1) Correction of the total elemental iodine removal coefficient for the sprayed region to 20.975 hr 1 , 2) Correction of the equilibrium core source term for Y90, Y91, Y92, Zr95, Zr97, Nb95, and Mo99, 3) Correction of the ESF leakage timing, and 4) removal of the unnecessary hydrogen mixing system model (sensitivity).

Revision 1 also incorporates Addendum A-01 and Post-It-Note P-01.

SUMMARY

OF RESULTS/CONCLUSIONS The post-accident offsite, Control Room, and Technical Support Center doses for a postulated design basis LOCA meet the requirements of 10 CFR Part 50.67. The LOCA dose results, including all leakage pathways, from Table 11-1, are given below: Location LOCA Dose remTEDE 21.2 6.9 3.0 0.5 LIMITATIONS OR RESTRICTIONS ON CALCULATION APPLICABILITY:

This calculation determines the radiological dose consequences resulting from the reader coolant release that accompanies a postulated design basis LOCA, which are reported in USAA Section 15.6.5. This calculation will become the licensing basis LOCA dose analysis after the transition to GNF2 fuel. IMPACT ON OUTPUT DOCUMENTS The results of this calculation will be incorporated into the USAR following LAR approval.

Page iv DOCUMENT INDEX 8 0 c z Revision, I! -s c J! :J i5 Document Numbermtle Edition, Date /!}_ D. c 1 Safety Evaluation by the Office of Nuclear Reactor Regulation Related to N/A 181 0 D Amendment No. 103 to Facility.Operating License No. NPF-58 2 NUREG/CR-6604, RADTRAD: A Simplified Model for Radionuclide Transport December 181 0 D and Removal and Dose Estimation 1997 3 NUREG-1465, Accident Source Terms for Light-Water Nuclear Power Plants February 181 0 D 1995 4 Federal Guidance Report 11, Limiting Values of Radionuclide Intake and Air Second 181 D D Concentration and Dose Conversion Factors for Inhalation, Submersion, and Printing, 1989 Ingestion 5 Federal Guidance Report 12, External Exposure to Radionuclides in Air, Water, September 181 0 D and Soil 1993 6 NUREG/CR-5966, A Simplified Model of Aerosol Removal by Containment June 1993 181 0 D Sprays 7 Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating July2000 181 D 0 Design Basis Accidents at Nuclear Power Reactors*

8 Not Used 0 0 D 9 GEH-KL 1WX23P-017, from E. G. Thacker II (GE) to E. S. Tomlinson Ill May7, 2012 D 181 D (FENOC), PNP GNF2 Fuel Transition:

F0802 Source Term Output Files 10 DES/98-0845, Telephone and Conference Memorandum by Paul J. 1212198 0 181 D Roney/DES, Perry Control Room Atmospheric Dispersion Factors (Chl/Q) 11 PSAT 150.01C.03, Dose Calculation Data Base for Application of the Revised Revlslon2 0 181 0 OBA Source Term to the Perry Power Uprate 12 PSAT 04202H.04, Aerosol Decay Rates (Lambda) in Drywall RevisionO 0 181 0 13 PSAT 04202U.03, Dose Calculation Data Base for Application of the Revised Revision2 D 181 0 OBA Source Term to the CEI Perry Nuclear Power Plant 14 10 Code of Federal Regulations 50.67, Accident Source Terms December 23, 181 0 D 1999 15 PNPP Technical Specifications Amendment 0 181 0 150 16 M26-001, M26, Volume Calculation, Control Room Envelope Revision2 0 181 0 17 PSAT 04202H.13, Offsite and Control Room Dose Calculation Revision 1 ti 181 0 18 PERRYUSAR Revision 17 181 0 D Pagev 8 0 c !? 'S z Revision, ! 'S c Document Number/T"itle Edition, Date a. i=i ..s 19 NEI 99-03, Control Room Habitability Assessment Guidance March2003 181 D D 20 Regulatory Guide 1.52, "Design, Inspection, and Testing Criteria for Air Revision2, 181 D D Filtration and Adsorption Units of Post-Accident Engineered-Safety-Feature March 1978 Abnosphere Cleanup In Light-Water-Cooled Nuclear Power Plants" 21 PNPP Technical Specification Section 5.5.7, PNPP Ventilation Filter Testing Amendment 181 D D Program (VFTP) 143 22 NUREG/CR-6604, Supplement 1, RADTRAD: A Simplified Model for June8, 1999 181 D D Radionuclide Transport and Removal and Dose Estimation 23 NUREG/CR-6604, Supplement 2, RADTRAD: A Simplified Model for October2002 181 D D Radionuclide Transport and Removal and Dose Estimation 24 10 Code of Federal Regulations 20, Standards For Protection Against Oct. 1, 2007 181 D D Radiation 25 PSAT 08401T.03, Peny Plant Total Effective Dose Equivalent (TEDE) Revision 5 D 181 D Calculation 26 PNPP Calculation 3.2.15.17, Containment Water Pool pH Post-Accident RevisionO D 181 D 27 GE letter from D. Braden (GE) to E. Root (CEI), GE-PAIP-651, ORF A22-NIA D 181 D 00084-00, dated 3/12101, *Additional Containment Response Cuives" 28 10 Code of Federal Regulations Part 50 -Domestic Licensing of Production and 17 FR 39906, 181 D D Utilization Facilities Jul. 6, 2012 29 10 Code of Federal Regulations Part 100-Reactor Site Criteria 77 FR39910, 181 D D Jul. 6, 2012 30 RADTRAD Computer Program Certification, FNOCPP184, CEl-120 181 D D 31 CEI Calculation 3.2.6.4, Post-LOCA Doses with Spray at 10 min for 6 Hours RevisionO 181 D D and Control Room lnleakage of 90 CFM 32 Wolfram Mathworfd httj!://mgthworfd.wglfram,com/Oblalg§Rheroid.!Jlml accessed 181 D D 712012012 33 NUREG/CR-0009, Technological Bases for Models of Spray Washout of 0 181 D D Airborne Contaminants in Containment Vessels 34 NUREG-800, Standard Review Plan (SRP) 6.5.2, Containment Spray As A Revlsion4, 181 D 0 Fission Product Cleanup System March2007 35 PNPP Drawing 320-0661-00000, Containment Spray System RevlsionT D 181 D 36 PNPP Drawing D-314-861, Sheet 3, Containment Vessel Spray Ring "A" Piping RevisionB D 181 D 37 PNPP Drawing D-314-661, Sheet 8, Containment Vessel Spray Ring *e* -RevlsionB D 181 D Page vi 8 0 c -z Revision, 2! :; :I c .!! & i5 Document Numberfrdle Edition, Date 6 Piping 38 PNPP drawing Ss-304-661, Sheet 105.2, Piping Isometric-Containment Spray RevisionC 0 181 0 System 39 PNPP Drawing D-314-661, Sheet 7, Containment Vessel Spray Ring "D" RevisionB 0 181 0 40 PNPP Drawing SS-304-661, Sheet 102.2, Piping Isometric-Containment RevlsionB 0 *181 D Spray System Reactor Building 41 PNPP Drawing D-314-661, Sheet 6, Containment Vessel Spray Ring "F" Revision B D 181 0 42 PSAT 04202H.08, Steamline:

Particulate Decontamination Calculation.

Revision 1 D 181 0 43 PSAT 04212H.02, Drywell Sweep-Out Rate and Related Thermal-Hydraulic Revision 1 181 D 0 Conditions Inside Containment 44 NUREG-0800, Standard Review Plan, 15.6.5, Appendix B, Radiological Revision 1, 181 D 0 Consequences of a Design Basis Accident:

Leakage from July 1981 Engineered Safety Feature Components Outside Containment 45 PNPP Drawing 511-0016-00000, Reactor Building -Steel Framing Sections Revision O D 181 0 and Details 46 CEI Calculation 3.2.6.3, LOCA Doses as a Fundlon of Spray Initiation lime Revision 0 181 0 0 I 47 003.008-001-00, FSAR Figure 3.8-1, "Typical Section of Reador Building Revision 12 181 0 0 Complex." 48 NUS letter from A. E. Mitchell (NUS) to R. F. Zucker (CEI), CSA-8106187-12, N/A D D 181 PY-NUS/CEl-1474, dated 312198, "Habitability Chi/Os for Tse*. Attached to PNPP Cale 5.7.1.2, page 4. I 49 PNPP Calculation

5. 7 .1.2, Technical Support Center -Final Dose Revision 0 D D 181 50 PNPP Drawing E-013-011, Final Plant Layout, Section A-A Revisions D 181 D 51 PNPP Drawing E-002-002, Final Plant Layout, Section A Revision 15 D 181 0 52 PNPP Drawing D-912-610, Control Room HVAC and Emergency Recirculation Revision FF D 181 0 System 53 WASH-1400, "Reactor Safety Study: An Assessment of Accident Risks In U.S. 1975 181 D 0 Commercial Nuclear Power Plants,* NUREG 75/014, Nuclear Regulatory Commission, Washington, DC. 54 D.I. Chanin, J.L. Sprung, L.T. Ritchie, and H-N Jow, "Melear Accident 1990 181 D 0 Consequence Code System (MACCS) User's Guide,* NUREG/CR-4691, Vol. 1, Sandia National Laboratories, Albuquen::iue, NM. 55 NUREG/CR-6189, DA Powers et al, "A Simplified Model of Aerosol Removal July 1996. 181 D 0 by Natural Processes in Reactor Containments." '5(o lf'lc.. T6.CUHIC.AL to BJ D 0 '") '"1*1'1 Page 1 1.0 PURPOSE The purpose of this calculation is to prepare a dose analysis supporting the transition to GNF2 fuel and to establish the new design basis loss-of-coolant accident (LOCA) dose analysis.

This calculation makes the current LOCA dose calculation (PSAT 08401T.03, DIN 25) and Technical Support Center calculation (Calculation

5. 7.1.2, DIN 49) historical and removes certain conservatisms contained in the current LOCA calculation to increase the margin of safety. This calculation will be performed in accordance with the guidance provided in Regulatory Guide 1.183 (DIN 7) for application of an alternative accident source term and will demonstrate that the offsite and onsite post-accident doses comply with the requirements and dose limits of 10 Code of Federal Regulations (CFR) Part 50.67 (DIN 14). Onsite doses calculated include the Technical Support Center (TSC) dose. The evaluation of the limiting design basis loss-of-coolant accident will use the RADTRAD 3.03 Code Instead of the proprietary STARDOSE code used in PSAT 08401T.03.

RADTRAD 3.03 was developed for the NRC and is commonly used in the nuclear power industry for applications of this type. Excess conservatisms removed from the current calculation (PSAT 08401T.03) are given In detail in the following sections and are summarized below:

Increased Control Room Emergency Recirculation System (CRERS) charcoal filter efficiency for elemental and organic iodine from 50% to 80%.

Credit for decay during the two (2) minute onset of the gap release.

Credit for elemental and aerosol removal in the unsprayed containment region.

Credit for reduced containment and annulus bypass leakage after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months based on accident containment pressure.

Increased CRERS HEPA filter efficiency from 95% to 99%. An additional conservatism added to this calculation is the removal of credit for auto-initiation of the CRERS. Isolation of the normal ventilation system and actuation of CRERS is assumed to be performed manually from the control room at 30 minutes post-accident.

2.0 BACKGROUND

The Perry Nudear Power Plant (PNPP) pilot Alternative Source Term (ASn submittal to the NRC was based on the LOCA analysis presented in PSAT 08401T.03, Revision 5 (DIN 25). This analysis utilized the POLESTAR proprietary computer code STARDOSE to determine the offslte and onsite consequences of a LOCA. The NRC, in approving the PNPP pilot Alternative Source Term (AST) submittal, performed a confirmatory radiological consequence calculation that evaluated potential fission product release pathways following a postulated LOCA. The NRC calculation was documented in the Perry Safety Evaluation by the Office of Nuclear Reactor Regulation related to Amendment No. 103 (DIN 1). The NRC staff used the RADTRAD Code.

Page2 The guidance of Regulatory Guide 1.183, uAltemative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors* (DIN 7), will be used to identify the conservatisms currently being applied in the Perry design basis LOCA model. Regulatory Guide (R.G.).1.183 establishes an acceptable Alternative Source Term (AST) and Identifies the significant attributes of other ASTs that may be found acceptable by the NRC staff. This guide also identifies acceptable radiological analysis assumptions for use In conjunction with the accepted AST. This calculation will remove some conservatism per the guidance of R.G. 1.183. 3.0 ACCEPTANCE CRITERIA The post-accident offsite and control room doses must meet the requirements of 10 CFR Part 50.67, u Accident Source Term.* 10 CFR 50.67 gives the limits applicable to plants revising their accident source terms. The dose limits specified are given in § 50.67. (b)(2)(i), (iO. and (iii) as follows: (b)(2Xi)-An individual located at any point on the boundary of the exclusion area for any 2-hour period following the onset of the postulated fission product release, would not receive a radiation dose in excess of 0.25 Sv (25 rem) total effective dose equivalent (TEDE). (bX2)(ii)

-An individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), would not receive a radiation dose in excess of0.25 Sv (25 rem) total effective dose equivalent (TEDE). (b)(2Xiii) -Adequate radiation protection is provided to permit access to and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of0.05 Sv (5 rem) total effective dose equivalent (TEDE) for the duration of the accident For plants Implementing the alternative radiological source term methodology, the dose limits of 10 CFR 50.67, given above, replace the limits given In 10 CFR 100.11, uDetermination of exclusion area, low population zone, and population center distance;*

which are expressed in terms of whole body and thyroid dose as follows: (a)(l) An exclusion area of such siz.e that an individual located at any point on its boundary for two hours immediately following onset of the postulated fission product release would not receive a total radiation dose to the whole body in excess of 25 rem or a total radiation dose in excess of 300 rem to the thyroid from iodine exposure. (aX2) A low population zone of such size that an individual located at any point on its outer boundary who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage) would not receive a total radiation dose to the whole body in excess of 25 rem or a total radiation dose in excess of 300 rem to the thyroid from iodine exposure.

Page3 As noted above, the dose limit for control room personnel Is specified In 10 CFR Part 50.67 (DIN14). 4.0 METHOD OF ANALYSIS This calculation will evaluate the total effective dose equivalent (TEDE) for the PNPP design basis radiological accident (LOCA) using the revised accident source term based on Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors* (DIN 7). The TEDE dose is defined as the sum of the deep-close equivalent (for external exposures) and the committed effective dose equivalent (for Internal exposures) (DIN 24). The RADTRAD Code, Version 3.03, will be used to calculate radiological consequences in terms of TEDE. RADTRAD <Radionuclide Iransport and Removal .@nd Dose Estimation) calculates fission product transport and removal along with the resulting radiation doses at selected receptors.

The code is described in NUREG/CR-6604, "A Simplified Model for Radionuclide Transport and Removal and Dose Estimation" (DIN 2, DIN 22, and DIN 23). RADTRAD 3.03 was certified for this application (DIN 30) in accordance with the ENERCON computer code certification procedure

[ENERCON CSP 3.02]. 5.0 ASSUMPTIONS 5.1 Control Room Emergency Reclrculatlon System (CRERS) Upon receipt of an Engineered Safety Feature (ESF) actuation system signal or high radiation signal, the PNPP control room heating, ventilation, and air conditioning (HVAC) system is designed to automatically isolate and activate the CRERS; however, this analysis conservatively assumes that the normal HVAC system continues to operate with an outside air intake (6000 cfm+10% margin) and exhaust to the environment (4800 cfm) until the CRERS is manually actuated at 30 minutes. Each redundant CRERS subsystem has a high efficiency particulate air (HEPA) filter, charcoal adsorbers and a post HEPA filter. Operation of the CRERS fans, charcoal adsorbers, and HEPA filters are credited in this analysis.

The CRERS Is an ESF system that Is tested in accordance with R.G. 1.52 (DIN 20). The current test acceptance criterion for the CRERS charcoal adsorbers requires a penetration of less than 2.5% (DIN 21). Based on the testing requirements, a charcoal adsorber removal efficiency of 95% could be justified; however, for additional operational margin, elemental and organic Iodine removal efficiency is assumed to be 80%. Technical Specification 5.5.7 (DIN 21) states that each HEPA filter is tested to show a penetration and system bypass of less than 0.05% when tested in accordance

with Regulatory Guide 1. 52 (DIN 20). A penetration and bypass of less than 0.05% allows credit for a particulate removal efficiency of 99% per Regulatory Position C.5.c of Regulatory Guide 1.52. This analysis uses a HEPA filter efficiency of 99 percent for aerosol particulates.

5.2 Hydrogen Mixing System The hydrogen mixing system is manually initiated.

For this analysis, operation of the hydrogen mixing system is not assumed. Due to the minimal (500 cfm) flow rate between the drywall and containment Page4 there is little effect on drywell or containment radionuclide concentrations due to operation of this system. Requiring operators to manually initiate the hydrogen mixing system early in the accident is not a good utilization of operator effort because of the minimal impact on accident doses. 5.3 Control Room lnleakage As described in Section 6.4 of the PNPP USAR (DIN 18), the control room is normally maintained at a slightly positive pressure to the surrounding areas from the 6600 cfm (includes a 10% tolerance on flow rate) fresh air makeup and out leakage of 4800 cfm. In the isolated mode, there is no intake from outside air sources and the control room pressure would eventually reach that of the surrounding areas. After the CRERS is initiated, the maximum control room unfiltered inleakage of 1375 cfm, will be used (DIN 11). The leakage out of the control room envelope is also modeled as 1375 cfm to avoid pressurization of the control room envelope.

5.4 Drywell Flows Leakage from the drywell into the primary containment is due to steaming from the heated reactor core in accordance with R.G. 1.183, Appendix A, Assumption

3. 7. This leakage is assumed during the two-hour period between the initial blowdown and termination of the fuel radioactivity release (gap and early vessel release phases). The termination of the release from the core is due to core recovery and reflood. Instead of evaluating all of the potential steaming rates due to various reflooding scenarios, this analysis will assume that there is a homogenous mixture in the drywell and containment starting at two hours. The assumption of a well-mixed drywall and containment atmosphere at two hours is appropriate because the EAB radiological doses consider the worst two hours as opposed to the first two hours as was done for the previous TIO 14844 source term methodology.

The assumption of a well-mixed drywell and containment atmosphere is implemented by assuming a high mixing flow (2. nE+05 dm, approximately one drywell volume per minute) between the two volumes. The mixing flow is conservatively assumed to continue for the duration of the accident instead of isolating the drywall after the core is quenched.

The flow rate from the Drywell to the Wetwell, which bypasses the suppression pool, is given in PSAT 150.01C.03 (DIN 11, page 6) as follows: Table 6-1 Drywall and Wetwell Mixing Flows Time After Gap Release Flow from OW to WW Flow from WW to DW (hours) (cfm) (cfm) 0-0.5 0 0 0.5-2.0 3000 0 Pages 2.0-720 2.77E+05 *2.77E+05 5.5 Containment Leakage Rate The primary containment consists of a drywell, a wetwell and supporting systems to limit fission product leakage during and following the postulated LOCA with rapid isolation of the containment boundary penetrations.

The secondary containment will collect and retain fission product leakage from the primary containment and will release fission products to the environment through the Annulus Exhaust Gas Treatment System (AEGTS). During normal operation, the shield building is maintained at a slight negative pressure.

Following a OBA, it is expected to remain negative, however for a short period it may not be maintained below the design negative pressure value of 0.25-inch.

water gauge (USAR 6.5.3.2.1).

Therefore, it will be assumed that the primary containment leakage is released directly to the environment for the first 40 seconds following the LOCA (DINs 11 and 25). However, because the gap release does not begin until two minutes post-accident, this 40-second period when there may be direct release to the environment is not considered.

5.6 AEGTS Filtration The AEGTS includes HEPA filters which are periodically tested to demonstrate compliance with Regulatory Guide 1.52. Particulate removal by the HEPA filters is assumed to be 99% in accordance with Regulatory Guide 1.52 (DIN 20). The system also contains 4-inch deep activated charcoal adsorbers to remove elemental and organic iodine; however, this analysis conservatively assumes a removal efficiency of 0% for the charcoal adsorbers to allow operational flexibility.

The AEGTS extracts and filters a maximum of 2000 cfm from .the annulus. During an accident, the maximum expected discharge to the atmosphere is 1000 cfm (DIN 18). The balance of the filtered AEGTS flow is routed back to the annulus. This analysis conservatively assumes that 2000 cfm Is discharged directly to the environment with no recirculation (holdup) of iodine in the annulus. 5.7 Containment Spray Manual initiation of containment spray is assumed at thirty minutes instead of automatic initiation on high pressures and low water level per previous (DIN 25) analysis.

Containment spray is assumed to end at 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at which time the radionuclide removal by containment spray is terminated.

USAR Subsection 6.5.2.3 gives a discussion of the non-mechanistic assumption that sprays will operate up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . In an actual event, spray use would not necessarily be suspended at 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months if appropriate conditions for their use still existed. Therefore, the assumption that sprays stop at 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is not intended to be interpreted as a commitment to stop using sprays after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . In addition, the statement that sprays will operate up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months implies that the sprays will not necessarily operate continually for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . The containment sprays will be run when it is appropriate, and not necessarily the entire time during the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of a LOCA. However, this does not invalidate the assumption used in this calculation.

The Pages accident guidance to operators is symptom based, rather than event based. Most postulated LOCAs will not result in large radiation releases and would not require containment sprays to run for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for removal of radioactivity from the containment.

Sprays are also used to reduce containment pressure, as needed, by steam condensation and containment heat removal. If a high radiation signal is present from the containment radiation monitor and pressures are elevated in containment, the sprays would be operated.

However, if containment gauge pressure is reduced to near zero and use of the sprays is terminated by the operators, this does not have an adverse impact on off-site doses (or the dose calculations) since the driving pressure for containment and main steam line leakage has been eliminated.

The dose calculations assume that the maximum allowable leakage (la) corresponding to the peak post-accident pressure (Pa) remains during this first 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period, so If containment pressure is reduced to substantially less than Pa, a reduction in leakage and the resultant offsite doses will follow. If containment pressure increases again, and the high radiation signal is present, sprays would be actuated again. 5.8 ECCS Leakage Consistent with the previous analysis (DIN 25), this analysis assumes that the ECCS leakage is 15 gallons per hour (gph) for the entire duration of the accident.

Additionally, leakage from a gross failure of a passive component is assumed to occur at a rate of 50 gallons per minute (gpm) starting 24 hour8 into the accident and lasting 30 minutes in accordance with NUREG-0800 SRP 15.6.5, Appendix B (DIN 44) : Regulatory Guide 1.183, Appendix A, Section 5.2 states that engineered safeguards feature leakage should be assumed to start at the earliest time that recirculation flow occurs in these systems and end at the latest time the releases from these systems are terminated.

For PNPP, ECCS leakage may begin up to 30 minutes post-accident but is assumed to begin at the onset of gap release at two minutes and continue for the duration of the event. This is a conservative assumption which maximizes the dose contribution for this release pathway. 5.9 l\'ISIV Leakage Rate There are four main steam lines; each line has an inboard MSIV, an outboard MSIV, and a third isolation (shutoff) valve. This analysis assumes a double guillotine pipe rupture in one of the four main steam lines upstream of the inboard MSIV and failure of all four third main steam shutoff valves (1N11-F0020A, B, C, and D) to close as a result of a common power failure (single-failure criterion).

A total of a 250 scfh maximum allowable leakage limit (TS SR 3.6.1.3.10) is assumed to occur: (1) 100 scfh through the broken steam line, (2) 100 scfh through a second intact steam line, and (3) the remaining 50 scfh through a third Intact steam line. This is modeled as 100 scfh through the broken steam line and 150 scfh through the unbroken steam lines. The calculation converts this to non-standard conditions, as explained in more detail in Section 6.12, "MSIV Flows:

Page7 5.1 O Bypass Leakage A 'PoRllo.-..

of Secondary containment bypass leakage is iR afifiitieft ta the containment allowable leakage, La. The leakage paths include all pathways which could potentially allow leakage to bypass secondary containment.

Therefore, any bypass leakage releases would not be treated by an ESF filtration system prior to being released to the environment.

The secondary containment bypass leakage is currently limited to 5.04% of La. when pressurized to by Technical Specification SR 3.6.1.3.9 (DIN 15) even though the previous LOCA dose calculation (DIN 25) assumed a leakage of 10.08% of L 8* The containment bypass leakage will be maintained at 0.1008 La in this analysis to allow for an increase In the Technical Specifieation allowable leakage limit. 5.11 Source Tenn Release In accordance with R.G. 1.183 (DIN 7), only the gap and in-vessel release phases are considered in this design basis LOCA dose calculation.

The core source terms are assumed to be released at a constant rate such that the release is completed by the end of the specified release period. Assumptions regarding release fractions and timing are consistent with Tables 1and4 of R.G. 1.183 (DIN 7). Table 1 of R.G. 1.183 Is given below: B'WR Core Inventory Fracdon Released Into Con&alnmeat GroDPl P1ii9Pbase Noble Gases 0.05 0.95 1.0 Halotzens 0.05 0.25 0.3 Alkali Metals 0.05 0.20 0.25 Telhui1101 Metals 0.00 0.05 0.05 Ba. Sr o.oo 0.02 o.oi Noble Metals 0.00 0.0025 0.0025 Ceri1un Group 0.00 0.0005 0.0005 Lanthanides o.oo 0.0002 0.0002 Table 4 of R.G. 1.183 Is reproduced below: LOCA Release Phases PWRs BWRs Phase Gap Release Early In-Vessel Onset 30 sec O.Shr Duration Onset 0.5 hr 2 min 1.3 hr 0.5 tu* Duration O.Shr 15hr 6.0 Design Input 6.1 Plant Grade The PNPP plant grade elevation is 620 feet (DIN 50). 6.2 Core Source Terms and Releases Pages The PNPP core source term release magnitude, timing and chemical form are based on Regulatory Guide 1.183 (DIN 7). The core source terms were developed by GE Hitachi (DIN 9). The calculated Inventories are based on 2-year GNF2 refueling cycles and serve as input for design basis accident analyses based on Regulatory Guide 1.183 source term assumptions.

The fission* product inventory calculations were performed using the ORIGEN2 code. The Ci/MN multipliers developed In DIN 9 are applied here to generate the core source terms at the onset of the event. A reactor power level of 3833 MWt will be used based on 102% of the rated thermal power, 3758 MWt, as defined in Technical Specification 1.1, Definitions, page 1.0-5. Amendment 112 (DIN 15). The GNF2 fuel source terms are based on the GNF2 equilibrium source activity given below. The source terms include fission products, actinides, and activation products.

The listing of isotopes given In Table 6-1, below, is based on the isotopes used in the RADTRAD computer code. As stated in the RADTRAD User's Manual, NUREG/CR-6604 (DIN 2), the 60 isotope nuclide file is based on isotopes selected in WASH-1400

[DIN 53) with the addition of 6 Isotopes used in the MACCS code [DIN 54). Co58 Co60 Kr85 Kr85m Kr87 Kr88 Rb86 Sr89 Sr90 Sr91 Sr92 Y90 Y91 Y92 Y93 Zr95 Table 6-1 Source Term GNF2 Equlllbrium lsotone fCllMWth) 2.647E+02 4.827E+02 3.789E+o2 6.737E+o3 1.283E+o4 1.804E+04 6.882E+01 2.425E+04 3.016E+03 3.064E+04 3.346E+04 3.118E+03 3.152E+04 3.362E+04 3.928E+04 4.440E+04 Isotope Zr97 Nb95 Mo99 Tc99m Ru103 Ru105 Ru106 Rh105 Sb127 Sb129 Te127 Te127m Te129 Te129m Te131m Te132 1131 1132 1133 1134 1135 Xe133 Xe135 Cs134 Cs136 Cs137 Ba139 Ba140 La140 La141 La142 Ce141 Ce143 Ce144 Pr143 Nd147 Np239 Pu238 Pu239 Pu240 GNF2 Equfllbrfum fCllMWth) 4.490E+o4 4.463E+o4 5.105E+04

4.470E+o4 4.309E+o4 3.046E+o4 1.750E+04 2.871E+o4 3.016E+03 8.906E+03 2.997E+03 4.049E+02 8.762E+03 1.304E+03 3.965E+03 3.850E+o4 2.714E+o4 3.914E+o4 5.495E+o4 6.025E+o4 5.150E+o4 5.302E+o4 1.934E+o4 6.926E+03 2.162E+03 4.190E+03 4.877E+04 4.709E+o4 5.002E+o4 4.440E+o4 4.278E+o4 4.460E+o4 4.090E+o4 3.670E+o4 3.957E+o4 1.795E+o4 5.619E+05 1.338E+02 1.291E+01 1.749E+01 Pages I 1* I Page10 GNF2 Equlllbrlum tsotoDe (CllMWth)

Pu241 5.748E+03 Am241 7.237E+OO Cm242 1.799E+03 Cm244 1.124E+02 6.2.1 Onset of Gap Release Table 4 of Regulatory Guide 1.183 tabulates values acceptable to the NRC for the onset and duration of each sequential release phase for OBA LOCAs at PWRs and BWRs. The specified onset of the gap release Is the time following the initiation of the accident (i.e., time = 0) prior to the start of the gap release. For a BWR the onset is 2 minutes. Credit will be taken for decay prior to the onset of the gap release at 2 minutes. 6.2.2 Release fractions The release fractions used in this analysis are consistent with Table 1 of R.G. 1.183 (DIN 7) which is reproduced in Section 5.11 of this calculation.

6.3 Suppression Pool Iodine Re-evolution The impact of any postulated iodine re-evolution from the suppression pool has been evaluated and shown to be negligible based on the pool pH level. If the pH is maintained above 7, very little (less than 1 %) of the dissolved Iodine will be converted to elemental iodine (DIN 1, DIN 7). The Standby Liquid Control System (SLCS) is used for controlling and maintaining long-term suppression pool water pH levels to 7 or above. The pH of post-accident water in the containment will remain above 7 for the entire duration of the postulated LOCA (DIN 26). As such, this analysis will not consider any impad to the offsite or control room doses due to iodine re-evolution from the suppression pool. Also, in accordance With Appendix A of Regulatory Guide 1.183 (DIN 7), because the suppression pool pH is controlled at values of 7 or greater, the chemical form of radloiodlne released to the containment can be assumed to be 95% cesium Iodide (Csl), 4.85 percent elemental iodine, and 0.15 percent organic iodide. 6.4 Dose Conversion Factors The effective dose conversion factors for the TEDE calculations are based on FGR 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion" (DIN 4) and FGR 12, "External Exposure to Radlonuclldes In Air, Water, and Soiin (DIN 5). These reports tabulate dose coefficients for external exposure to photons and electrons emitted by radionuclides distributed in air, water, and soil, as well as, dose coefficients for the committed dose equivalent to tissues of the body per unit activity of inhaled or ingested radlonuclides.

These dose coefficients for exposure to radiation are intended for the use in calculating the dose equivalent to organs and tissues of the body and are endorsed by the NRC in Regulatory Guide 1.183, Sections 4.1.2 (FGR Page 11 11) and 4.1.4 (FGR 12). Dose conversion factors for the 60-isotope, 9 element NUREG 1465 {DIN 3) accident source term composition are included in the RADTRAD Input. 6.5 Atmospheric Dispersion Factors The atmospheric dispersion factors ('xjQ values) for the LPZ and EAB are obtained from PSAT 04202U.03 (DIN 13, page 10) and Calculation 3.2.6.3 (DIN 46). The Control Room atmospheric dispersion factors are documented in DES/98-845 (DIN 10). The Technical Support Center atmospheric dispersion factors are documented in PY-NUS/CEl-1474 (DIN 48) and Revision 0 of the TSC Dose Evaluation (DIN 49). The x/Q values, based on a ground level release, are given below: Table&-3 'IJQ (seclm 3) Location*

Time Interval EAB LPZ Oto 2 hrs 4.3E-4 4.8E-5 2to8hrs 4.8E-5 8to 24 hrs 3.3E-5 24to 96 hrs 1.4E-5 96to 720 hrs 4.1E-6 Table 6-4 Control Room and TSC 'IJQ (sec/m 3) Time Interval CONTROL ROOM TSC Oto 8 hrs 3.5E-4 5.1E-5 8 to 24 hrs 2.1E-4 4.1E-5 24 to 96 hrs 1.1E-4 3.1E-5 96to720 hrs 5.75E-5 1.1E-5 6.6 Breathing Rate and Occupancy Factors The breathing rates applied in the calculation of the Inhalation dose are consistent with those reported In Sections 4.1.3 and 4.2.6 of R.G. 1.183 (DIN 7). Time Period Oto 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Bto 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 1to30 days . ' Table&-5 Breathing Rates (m 3/s) EAB LPZ 3.5E-4 3.5E-4 1.BE-4 1.8E-4 2.3E-4 2.3E-4 Control Room/TSC 3.5E-4 3.5E-4 3.5E-4 Page 12 The. control room and TSC occupancy factor$ are consistent with those reported in Section 4.2.6 of R.G. 1.183 and are tabulated below.

Table 6-6 Control Room and TSC Occupancy Factors Time Period Occupancy Factor Oto 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 1.0 1 to4 days 0.6 4to30 days 0.4 6. 7 Containment Volumes The volumes of the containment regions are from CEI Calculation 3.2.6.4, Revision 0, Page 3A of 33 (DIN 31). Table 6*7 Containment Volumes Region Volume (ft"') Unsprayed Containment 684,226 Sprayed Containment 481,174 Drywall Note: the above volumes are shown as rounded values in the RADTRAD screen views but the actual values are used in the RADTRAD input file. 6.8 Technical Support Center Doses Doses to personnel in the Technical Support Center (TSC) are calculated in the same manner as the doses to the Control Room operators except for the TSC specific atmospheric dispersion, xfQ, values and TSC data. The additional data needed to determine the TSC doses is as follows (DIN 49, page 9 and 10, and DI 5.7.1 page 18): Parameter TSC Volume (ff) HVAC Flow Rate (cfm) Table&-8 TSCData Recirculation Filter Flow Rate (cfm) Charcoal Filter Bed Depth (in) Filtered Damper lnleakage (cfm)* Unfiltered lnleakage (cfm)** *Added to unfiltered inleakage Value 113,412 37,000 6,000 2 12 27.2 Page 13 -After recirculation is initiated at 60 minutes (includes 1 O cfm for ingress and egress) For the TSC charcoal removal efficiency, a removal efficiency of 80% will be used to provide margin as was done for the Control Room charcoal adsorber removal efficiency.

Note that in the RADTRAD files, the TSC is labeled as the Control Room. The normal HVAC flow is assumed to operate for the first 60 minutes after which It is isolated and the recirculation filter is initiated.

6.9 Mixing Between the Unsprayed Containment and Sprayed The mixing rate between the unsprayed containment and the sprayed containment was determined to be 71,400 dm in Calculation PSAT 04202U.03, Rev. 0 (DIN 13, page 6). 6.1 O Containment Leakrate The maximum allowable primary containment leakrate, La, is 0.2 volume percent per day at the peak containment pressure (Pa) of7.80 psig per Technical Specification 5.5.12 (DIN 15). Per Assumption 5.5, primary containment leakage is released directly to the environment for the first 40 seconds following the LOCA, when the shield building may not be at a negativ_e pressure.

This potential 40 second release directly to the environment is not modeled because it has no dose significance.

Following this 40-second period, the annulus will collect and retain any fission produd leakage from the primary containment and will release fission products to the environment through the AEGTS. Secondary Containment Bypass leakage is a portion of the total containment leakage, La. Technical Specification SR 3.6.1.3.9 (DIN 15) limits the secondary containment bypass leakage to equal to or less than 5.04 percent of the primary containment leak rate. The containment bypass leakage for this calculation Is assumed to be 0.1008 La. (Assumption 5.10) Therefore, the leakrate for the sprayed and unsprayed (including drywell) regions of the containment is calculated below: * [(4.812e + 05 ft 3)

0.2%] Leakrate from Sprayed Region = min 24hr

60liT Leakrate from Sprayed Region = 0.668 cfm Subtracting the bypass leakage of 0.1008*La (i.e., 0.067 dm) gives: Leakrate from Sprayed to Annulus = 0.668 -0.067 cfm = 0.601 cfm [(6.842e + 05 ft 3 + 2.765e + 05 ft 3)

0.2%] Leakrate f ram Unsprayed Regions = min 24hr*60rr Page 14 Leakrate from Unsprayed Regions = 1.334 cfm Subtracting the bypass leakage of 0.1008*La (i.e., 0.134 cfm) gives: Leakrate from Unsprayed Regions to Annulus = 1.334 -0.134 cfm = .. 1.2 cfm Bypass Leakage: Sprayed Region Bypass Leakage = 0.668 cf m

0.1008 = 0.067 cf m Usprayed Regions Bypass Leakage= 1.334 cfm

0.1008 = 0.134 cfm Total Bypass Leakage = 0.202 cfm Table6-9 Containment Leakrate Summary Leakrate (W/mln) Leakage 'IO sec. to 24 hrs From Sprayed Region to Annulus 0.601 From Unsprayed Regions to Annulus 1.2 Bypass Sprayed Region to Environment 0.067 Bypass Unsprayed Regions to Environment 0.135 (rounded up) 6.11 Leakage after 24 Hours Containment leakage depends upon containment pressure and will be reduced at 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months as allowed by Regulatory Gulde 1.183, Appendix A, Section 3. 7. Based on the post-accident containment pressure curve for a MSLB (DIN 27), the containment pressure at 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months post-accident is 18.1 psia. This value was obtained by digitizing the containment pressure curve and finding the pressure at 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Atmospheric pressure at the elevation of the PNPP site (620 ft AMSL) is 14.37 psia. Because the flow rate is proportional to the square root of the differential pressure, the reduction in flow rate may be estimated as follows (assuming all other parameters remain constant): =0.691=>0.69 7.8 Page 15 Because the secondary containment bypass leakage also depends upon containment pressure, this leakage rate will also be reduced to 69% of the pre-24 hour values beginning 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months post-accident.

Therefore, this analysis will reduce all leakage flows to 69% of the Df!f-24 hour value after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Table 6-10 <p1,j. Jw tJ./, *t'( Containment Leakrate after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Leakage Leakrate (tr/min) From Sprayed Region to Annulus 0.415 From Unsprayed Regions to Annulus 0.828 Bypass Sprayed Region to Environment 0.0462 Bypass Unsprayed Regions to Environment 0.0932 6.12 MSIV Flows The flows given in Section 5.9 are based on MSIV leakage rate testing requirements at standard conditions.

The drywell atmosphere will not be at standard conditions after the reactor blowdown.

Calculation PSAT 04202H-04 (DIN 12, page A3) determined that the total MSIV flow rate from 0 to 7484 seconds was 298 cfh based on the minimum post-accident drywell pressure of 15. 7 psia and minimum temperature of 215°F. From 7484 to 86400 seconds, the MSIV flow rate Is 247 cfh based on the minimum pressure of 15. 7 psia and temperature of 100°F. Based on the assumed flow split given in Section 5.9, the flow through the broken steam line is: 298 cf h 100 scf h Qbroken uneCt S 7484 seconds) = 60 min/ hr* 250 scfh = 1.987 cfm 247 cfh 100 scfh Qbroken uneC7484 < t :5 86400 seconds) = 60 min/ hr* 250 scfh = 1.647 cfm The flow through the intact steam lines (100 scfh through a second intact steam line and the remaining 50 scfh through a third Intact steam line) is given below.

298 cfh 150 scfh Qintactlines(t S 7484 seconds) = 60 min/ hr* 250 scfh = 2.98 cfm 247 cfh 150 scfh QfntactUnesC7484

< t S 86400 seconds)=

60 min/hr* 250 scfh = 2.47 cfm Page 16 Consistent with Section 6.2 of Appendix A to R. G. 1.183, this leak rate may be reduced by as much as a factor of 2 after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , if supported by plant specific analysis.

As calculated above, this analysis will reduce all initial leakage flows to 69% of the pre-24 hour. leak rate after 24 hrs. This leakrate is conservatively based on the Initial flows (i.e., t<7484 seconds).

The MSIV leakage flows at this time become: Table 6-11 Steam Line Leakrate after 24 Hours Leakage Path Leakrate (ff/min) Broken Steam Line 1.371 Intact Steam Lines 2.056 The main steam line leakage rate is required to be less than or equal to 100 scfh (Technical Specification SR 3.6.1.3.10) when tested at Pa (7 .80 psig). The test temperature is assumed to be 70°F and the RCS operating temperature is 552°F (saturated temperature at steam dome pressure of 1045 psig, LCO 3.4.12). Converting the 100 scfh leakage rate to operating temperature gives a leakage rate of 191 cfh. The MSIV flow after the first piping segment from the drywell will be based on a constant maximum steam line flow of 191 dh (3.183 cfm) (DIN 11, page 8) for the broken steam line and 191 cfh*150 scfh/100 scfh = 4. 775 cfm for the unbroken lines. No reduction in flow at 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for these *line segments will be taken. This conservative model is unchanged from the previous analysis (DIN 25). 6.13 Radionuclide Removal Mechanisms Removal mechanisms for elemental iodine and aerosols will be applied in this calculation using NRC correlations incorporated into the RADTRAD 3.03 code. 6.13.1 Removal by Deposition Elemental Iodine Removal by Deposition Elemental iodine removal is credited in the drywell and containment volumes. Airborne elemental iodine is removed by deposition to the walls in the drywell and containment.

As reported in Section 5.1.2 of NUREG/CR-0009, DIN 33), this process is driven by the temperature differences between the surfaces and the atmosphere.

The removal factor reported in NUREG/CR-0009 is given by the following equation . where: .ii.= K 9 A v 1i. = removal rate constant due to surface deposition, kg= average mass transfer coefficient 0.137 cm/s (16.18 ft/hr) from page 17 of NUREG/CR-0009, A = surface area for wall deposition, and V = volume of contained gas. Page 17 This formula is also reported in Standard Review Plan 6:5.2 (DIN 34) as a method of calculating the total elemental iodine removal capability.

These removal constants are applied until a decontamination factor {OF) of 200 has been obtained.

Volume and Area Calculations Drvwell Volume For all volume calculations, surfaces other than the inner and outer building wall will be conservatively neglected.

The PNPP drywell volume of 276,500 ft3 from CEI Calculation 3.2.6.4, Revision 0, Page 3A of 33 (DIN 31) is used in this calculation.

Wall Surface Area Considering the 36'6" inside radius (DIN 45) of the drywell cylinder and the approximately 66 foot height above the suppression pool high water level (DIN 51), the area of the inner drywall wall is calculated to be 16.000 ft2. The use of the suppression pool high water level is conservative because it minimizes the wall surface area available for deposition.

Area = rrDh = rr

73'

66' = 15136.2 f t 2 or, 15,000 /t 2 Sprayed Containment Region Volume Although in some parts of the containment, the containment spray would fall directly to the suppression pool, the refueling floor {grating) at El. 689'-6" would affect a large fraction of the containment spray. As such, the only containment volume credited with spray removal is that area above the refueling floor. The upper containment (sprayed) region volume of 481, 174 ft 3 from CEI Calculation 3.2.6.4, Revision 0, Page 3A of 33 (DIN 31) is used in this calculation.

Wall Surface Area The surface area is taken as the containment wall area above the refueling floor at 689'-6" (DIN 47) and below the containment spring line at 727' (DIN 47). Using the containment radius of 60' (DIN 45) the surface area is calculated below as 14, 137 ft2. Area = rrDh = rr

2

60' * (727' -689.5') = 14,137 /t 2 The surface area of the oblate elliptical spheroid above the spring line is given by: (DIN 32) c 2 * (1 + e) S = 2na 2 + 11-ln e 1-e Where "a" is the equatorial radius and *c" is the polar radius and the ellipticity, "e" is given by: e=V Substituting 60' for "a" and (757' -727' = 30') for uc" {DIN 47) gives: e= (30')2 1-(60')2 = 0.866 -* , 2 (30')2 (1 + 0.866) -2 S -2n (60) + 11 0_866 ln 1_0_866 -31,218 ft The surface area of the dome is half of this total area: S = 31,218 ft!-12 = 15,609 The total area available for plateout is therefore 29,746 ft!-or, 29,000 ft!-. Unsprayed Containment Region Volume Page18 The volume of the unsprayed containment region is 684,226 ft3 per from CEI Calculation 3.2.6.4, Revision 0, Page 3A of 33 {DIN 31). Wall Surface Area Considering the 41 '6" outside radius of the drywall {DIN 45) and the approximately 96 foot height (689' -6" -593'-4" = 96.17') above the suppression pool high water level {DIN 51), the area of the outside drywall wall is calculated to be 25,000 ft2. The use of the suppression pool high water level is conservative because it minimizes the wall surface area available for deposition.

Page 19 Area= nDh = 7r

2

41.5'

96' = 25,032 ft 2 or, 25,000 /t 2* The radius of the unsprayed containment wall is 60' gMng a surface area of Area = nDh = n

2

60'

96' = 36,191 f t 2 or; 36,ooo /t 2 This gives a total surface area of 61,000 ft3 Using the above wall areas and volumes, the removal rate constants are given below: Table 6-12 Elemental Iodine Deposition Removal Factors Removal Volume Wall Area Factor Node (tt3) (ft2) (h1"1) Drywell 276,500 15,000 0.878 Sprayed Containment 481,174 29,000 0.975 Unsprayed Containment 684,226 61,000 1.443 Airborne elemental iodine removal by deposition to the walls In the drywell and containment is assumed to end when a OF of 200 is reached. Aerosol (particulate)

Removal by Deoosltlon Regulatory Guide 1.183, Appendix A, Section 3.2 (DIN 7), discusses the reduction in airborne radioactivity in the containment by natural deposition.

This section references the model in NUREG/CR-6189 (DIN 55) as an acceptable model. This model (the "Powers* model) Is Incorporated Into the RADTRAD code. Aerosol removal in the drywell and unsprayed containment region Is based on the 10% Powers Aerosol model in RADTRAD. Note that, for unsprayed regions, the reactor and accident type used in the Powers aerosol model must be reset to *eWR-Deslgn Basis Accident" prior to each execution of the RADTRAD code. 6.13.2 Removal by Sprays Aerosol Removal by Sprays A simplified model for estimating the fission product aerosol removal by containment sprays following a postulated LOCA is used. The model for aerosol removal by sprays built into the RADTRAD 3.03 code is the Powers model. The model was developed using values of 10, 100, and 2500 cm 3 H 2 0/ cm 2-s for the spray water flux. The model should not be used for spray water fluxes and fall heights outside of these Page20 ranges (DIN 2, 22, and 23). The Powers model was derived by correlating the results of Monte Carto uncertainty sampling analyses assessing the uncertainties in aerosol properties, aerosol behavior, spray droplet behavior, and the Initial and boundary conditions expected to be associated with a postulated LOCA in the containment.

The Powers mechanistic model requires that the user specify the following:

1. Q, the spray water flux, in cfmlsq ft; 2. H, the fall height, in meters: 3. ALPHA, the ratio of unsprayed volume to sprayed volume, 4. PCT, the uncertainty percentile selected for the model (10th, Soth, 90th percentiles).

The spray alpha is 6.8423E+05/4.8117E+05

= 1.422. The other two parameters used in this evaluation that are not treated as uncertainty distributions for Perry are (1) spray water flux, and (2) mean spray fall height. These parameters are specified based on plant specific design information.

The ubest estimate" value is associated with the 50th percentile; or median values: the lower bound Is associated with the 10th percentile; and the reasonable upper bound, or largest decontamination factor (DF), with the 90th percentile.

For aerosol removal by containment spray, the RADTRAD Powers Model 10th percentile uncertainty distribution for fission product in aerosol form is used in this analysis.

Note that the reactor and accident type used in the Powers aerosol model must be reset to "BWR-Design Basis Accidenr prior to each execution of the RADTRAD code. The PNPP LOCA dose analysis credits spray removal of aerosols in the sprayed region of the containment.

The Powers spray removal model implemented in RADTRAD requires the spray flux and spray height as inputs. The spray flow is 5250 gpm (D-302-0661-00000, Rev. G, DIN 35) per train. Technical Specification 3.6.1. 7 requires that the spray flow from the RHR system be gpm. Because the Powers model spray removal is a direct function of spray flow rate, a lower bound of the spray flow (i.e., 5250 gpm) Is conservative. . _ Spray Flow _ 5250 gpm

0.1337 cf m/ gpm _ 620 cfm/ Spray Flux -Sprayed Area -n * (60')2 -O.O 6 sq ft The average droplet fall height is dependent on the available train of containment spray. As shown below, the headers for the "N Train are located above the headers for the "B" Train per* drawing D-320-0661-00000, Rev. G. If the flow rate through all nozzles is assumed to be equal, the average drop height can be calculated by the nozzle-weighted average of the drop heights. The average drop height is used because the train operating post-accident is unknown. The drop height is based on the distance above the operating floor at El. 689'-6" (DIN 51).

Page21 Table 6-13 PNPP Containment Spray Heights RHR Header Header Refe19nce Drawing Height* H1 Number of Ni*HI Train Designation Elevation (ft) Nozzles 7 -N 1 (ft) A A 735.250 D-314-661, Sheet 3, 45.75 129 5901.75 Rev. 8. fDIN 36l c 744.250 SS-304-661, Sheet 54.75 113 6186.75 105.2, Rev. C, (DIN 38) E 750.500 SS-304-661, Sheet 61.00 102 6222 102.2, Rev. B, (DIN 40) B B 737.000 0-314-661, Sheet 8, 47.50 129 6127.5 Rev. B, tDIN 37l D 745.750 0-314-661, Sheet7, 56.25 113 6356.25 Rev. B. COIN 39) F 752.000 D-314-e61, Sheet 6, 62.50 104 6500 Rev. B, COIN No. 41) ,_T.;..;o-.,;ta-.,1

__ ._..;6;;...;.9.;;..0

____ 37294.25 I Average {ft) 154.05 I 7 320-0661-00000, Rev. G (DIN 35) where: N 1 Is the number of nozzles on header I His the height of header i above the operating floor (ft) The average fall height for both trains combined is therefore 64.06 ft. As discussed in SRP 6.5.2 (DIN 34), because the removal of particulate material depends markedly upon the relative sizes of the particles and the spray drops, the aerosol spray removal lambda is assumed to decrease by a fader of 10 after the aerosol mass has been depleted by a fader *af 50.

Page22 Elemental Iodine Removal bv Spravs SRP 6.5.2 provides guidance on calculating the spray lambda for removal of elemental iodine. The following formula is valid for lambdas greater than 10 per hour with a maximum of 20 per hour to prevent extrapolation beyond the existing data. where: As= first-order removal coefficient by spray, kg = the gas-phase mass-transfer coefficient, T = the fall time of the drops, which may be estimated by the ratio of the average fall height to the terminal velocity of the mass-mean drop, F = volume flow rate of the spray pump, V = containment building net free volume, and D = mass-mean diameter of the spray drops. Gas Phase Mass Transfer Coefficient The gas-phase mass-transfer coefficient, kg, can be determined by back-calculation from a solved case with slightly different assumptions.

Specifically, the example on Page 106 of NUREG/CR-0009, "Technological Bases for Models of Spray Washout of Airborne Contaminants in Containment Vesselsn, 1978, (DIN 33) uses the stagnant film model to determine the spray removal coefficient for a PWR case with a 1713 spray nozzle and the following parameters.

As= 14.2 hr" 1 F = 1500gpm V = 1.75E6 ft3 Height = 90 ft Temp= 250°F Solving the above equation for kg, gives: A 5*V*D kg= 6*T*F To calculate kg, the values of the mass-mean drop diameter, D, and the fall time of the drops, T, are needed. The PNPP spray nozzles are Spraco 1713A nozzles (DIN 35). Recent test results with the Spraco 1713A nozzles presented in Figure 4 of NUREG/CR-5966 (DIN 6) have a mean droplet size of 234 1Jm (NUREG/CR-5966, page 7) and an upper diameter of about 1500 1Jm. The mass-weighted average drop size, however, will be larger than 234 microns since the larger drops have exponentially more mass. This volume-weighted size distribution (which is directly related to the mass-weighted distribution) is reported in Figure 7 of NUREG/CR-5966 which Illustrates an average of the volume Page23 weighted distribution to be approximately 1200 microns. A value of 1200 microns will be applied in this calculation.

The terminal velocity of 1200 µm drops can be found to be approximately 400 cm/s from Figure 16 of NUREG/CR-5966.

Conservatively assuming the velocity Is equal to the tenninal velocity, a 90 foot (27 43 cm) fall height gives a fall time of 6.86 seconds. Using the above data to determine the gas-phase mass-transfer coefficient, kg, gives: As* v

D m

loocm/m cm kg = 6

T

F = _ 90 ft cm ft: 3 = 6 sec 6

400 cm/sec* 30.4Bir

1500 gpm

0.1337 iii For PNPP, the average fall height of the spray drops is calculated to be 54.05 ft (1647 cm). The terminal velocity of 1200 tJm drops can be found to be approximately 400 emfs from Figure 16 of NUREG/CR-5966. The drop fall time is calculated to be 4.1 seconds. The spray flow is 5250 gpm from D-302-0661-00000, Rev. G, (DIN 35) and the sprayed volume of the containment is 481, 174 fl3 from CEI Calculation 3.2.6.4, Revision 0, page 3A of 33 (DIN 31). From the SRP equation, below, the PNPP spray lambda for elemental iodine can be calculated to be 107.66 hr. cm ft: 3 min 6 *kg

T

F 6

6sec

4.1sec*5250 gpm

0.1337 gal* 60nr _ _1 .ils = V

D = 481,174 ft:3

1200

l0-6 m .10ocm/m -107*66 hr This result is reasonable considering the 14.2 hr 1 value calculated for the PWR case described in NUREG/CR-0009, the much higher spray flow rate at PNPP, and the smaller sprayed volume at PNPP. Since the SRP allows a maximum lambda of 20 hr" 1 , this calculation will apply a spray removal lambda of 20 hr" 1 for elemental iodine. As discussed previously, elemental iodine is removed by deposition to the walls in the containment with a removal coefficient of 0.975 hr" 1 for the sprayed region which gives a total elemental iodine removal coefficient for the sprayed region of containment as 20.975 hr 1* As discussed in SRP 6.5.2, the maximum decontamination factor is 200 for elemental iodine. The effectiveness of the spray in removing elemental iodine will be presumed to end at that time, post-LOCA, When the maximum elemental iodine OF is reached.

Page24 6.14 Annulus Model The Annulus Exhaust Gas Treatment System (AEGTS) Is an engineered safety features system designed to collect, process, and release the fission product leakage from the primary containment into the shield building.

The system is operated continuously during normal operation and maintains a slight negative pressure In the shield building.

The AEGTS is a redundant system consisting of pre-HEPA filters, charcoal adsorbers and post-HEPA filters.

in release activity by ESF ventilation filtration systems may be credited where applicable if filter systems used in these applications are evaluated against the guidance of Regulatory Guide 1.52 (DIN 20). The AEGTS charcoal adsorbers are not credited for reducing the released activity, so testing in accordance with R.G. 1.52 is not necessary.

The AEGTS HEPA filter is tested in accordance with Regulatory Guide 1.52 to verify a penetration and system bypass of less than 0.05% (DIN 21). Aerosol removal by the HEPA filters is therefore assumed to be 99%. As discussed previously, no credit for charcoal filtration of the annulus exhaust is taken in this . calculation.

6.15 Deposition In Main Steam Lines The deposition in the main steam lines will use the aerosol removal efficiencies from PSA T 08401T.03 (DIN 25) which was based on PSAT 04202H.08 (DIN 42). These removal efficiencies include a 10% increase in aerosol penetration to add conservatism to the main steam line leakage pathway. The removal efficiencies are given below. Table 6-14 Main Steam Line Removal Fractions Time after MSL 1 MSL2 MSL3 release (hr) (failed steamline) (pipe to intact steamlines) (intact steamlines) 0.0 0.681 0.7206 0.714 0.5 0.835 0.7206 0.813 1.5 0.8713 0.7206 0.8361 3.0 0.89 0.7206 0.8449 5.0 0.8614 0.7206 0.8339 7.0 0.8185 0.7206 0.7998 9.0 0.769 0.7206 0.7558 11.0 0.3653 0.7206 0.3807 720 0.0 0.0 0.0 The elemental iodine removal efficiency is 0.45 for all steam lines (DIN 25).

Page25 7.0 ACCIDENT SCENARIO AND CHRONOLOGY 0 minutes to 2 minutes A design basis double-ended guillotine break occurs in a main steam line upstream of the inboard MSIV, releasing reactor coolant into the drywall. The drywell ls pressurized driving drywell atmosphere out the MSIVs and into containment via the drywell bypass. All MSIV and containment leakage is Initially directed to the environment.

The AEGTS system achieves a 0.25-inch vacuum in the secondary containment at 40 seconds (Assumption 5.5) and draws 2000 cfm of secondary containment atmosphere through a HEPA filter and charcoal bed before release to the environment.

Because there Is no core damage during this 40 second drawdown period, it is not included In the RADTRAD model. Following this drawdown period, all primary containment leakage is directed to the secondary containment except for the containment bypass leakage, which _is assumed to bypass secondary containment and is released directly to the environment.

No credit for elemental or organic iodine removal by the AEGTS charcoal adsorbers is taken. Particulate removal by the HEPA filters is assumed to be 99% in accordance with Regulatory Guide 1.52. The control room and offsite dose points begin to accumulate dose from the ECCS, MSIV and containment leakage. The control room normal ventilation mode is assumed to continue until the CRERS is manually initiated at 30 minutes. 2 minutes to 32 minutes

The gap release begins by releasing the gap source terms into the drywall at a constant rate over the 30-minute release period following the onset of gap release at two minutes post-accident ECCS leakage is assumed to begin at this time leaking contaminated suppression pool water (10% of iodine -all forms) directly to the environment even though the postulated core damage is occurring because no ECCS injeciion Is assumed to be available during the first two hours. The control room and offslte dose points begin to accumulate dose from the ECCS, MSIV and containment leakage. At 30 minutes post-accident, the control room normal ventilation system is manually isolated, and the CRERS is manually initiated.

The CRERS fans recycle 27,000 cfm of control room atmosphere through HEPA filters and charcoal adsorbers.

Manual initiation of containment spray is assumed at thirty minutes. Manual initiation of containment spray at 30 minutes is reasonable based on Emergency Operating Procedure guidance f D* requiring operation of containment spray based on the "Pressure Suppression Pressure" curve contained in the EOP. The containment pressure threshold is met within 30 seconds of the LOCA per DIN 27. 32 minutes to 122 minutes At 30 minutes, the control room normal ventilation system is manually Isolated, and the CRERS is manually initiated.

The CRERS fans recycle 27,000 cfm of control room atmosphere through HEPA filters and charcoal adsorbers.

The in-vessel release begins at 30 minutes after the onset of gap release Page26 by releasing the in-vessel source terms into the drywall at a constant rate over a 90-minute release period. Manual* initiation of containment spray is assumed at thirty minutes. Manual Initiation of containment spray at 30 minutes is reasonable based on Emergency Operating Procedure guidance requiring of containment spray based on the "Pressure Suppression Pressure*

curve contained in the EOP. The containment pressure threshold is met within 30 seconds of the LOCA per DIN 27. 122 minutes to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months The source term release from the vessel is terminated at 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after the onset of gap release with the actuation of ECCS, which results in large amounts of steam evolution and large flows out of the drywall into the containment.

The drywall and lower containment region are assumed to become well-mixed at 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ./ ,A,. ,.1.1*1'f

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to 30 days Releases to the environment via the containment bypass, MSIV leakage, ECCS leakage and AEGTS exhaust continue for 30 days. As discussed above, containment leakage is reduced at 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months based on containment pressure.

8.0 MODEL DEVELOPMENT This analysis considers the following three pathways through which source terms can be released from the containment

. ECCS liquid leakage outside of containment

MSIV leakage

Containment airborne leakage (containment bypass and containment leakage) These three pathways are discussed below. RADTRAD modeling capabilities allow incorporation of the MSIV leakage and the containment airbome leakage into one model, therefore; the three release pathways are addressed In two RADTRAD models. 8.1 ECCS Liquid Leakage 8.1.1 Source Tenns The gap and core activity is released to the drywell atmosphere based on the release fractions and timing reported in Tables 1 and 4 of R. G. 1.183 and is assumed to be immediately dissolved in the suppression pool. Only halogens are modeled In this analysis.

Noble gases are not soluble and, with the exception of Iodine, all other radioactive materials in the recirculating liquid should be assumed to be retained in the liquid phase. This is consistent with the guidance of R.G. 1.183, Appendix A.

Page27 8.1.2 Volumes The suppression pool inventory expected during the LOCA is 114,379 ft3 (DIN 11). No credit Is taken for holdup in the Auxiliary Building where the ECCS systems are located. 8.1.3 Flows The earliest that the containment spray system could potentially be automatically initiated to spray the Containment is 10 minutes post-accident if high containment pressure combined with other LOCA signals is sensed. For this calculation it is conservatively assumed that the ECCS system leakage begins immediately after the LOCA (at the beginning of the gap release at two minutes post-accident).

Consistent with the previous analysis (DIN 25), this analysis assumes that the ECCS leakage is 15 gph (0.0334 cfm) for the entire duration of the accident.

This Is twice the established administrative limit of 7.5 gph. Additionally, *leakage from a gross failure of a passive component is assumed to occur at a rate of 50 gpm (6.68 cfnl) starting 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months into the accident and lasting 30 minutes in accordance with NUREG-0800 (DIN 44). 8.1.4 Removal Mechanisms Because the suppression pool temperature will not exceed 212°F (DIN 11) during the accident, ten percent of the iodine in the ECCS leakage is assumed to become airbome consistent with R.G. 1.183, Appendix A. Natural removal mechanisms and holdup In the auxiliary building are conservatively neglected.

Consistent with Section 5.6 of R.G. 1.183, Appendix A, the chemical species of these airborne source terms Is assumed to be 97% elemental and 3% organic. 8.1.5 Model The ECCS liquid leakage model is Illustrated in Figure 1. 8.1.6 Results The radiological doses for the ECCS liquid leakage transport path are reported in Table 11-1. The RADTRAD output file, Including the input summary, is listed in Attachment

1. 8.2 MSIV Leakage 8.2.1 Source Tenns As discussed previously, the PNPP core source terms have been developed with the ORIGEN2 methodology.

These source terms are released into the drywell based on the release fractions and timing reported in Tables 1 and 4 of R.G. 1.183.

Page28 8.2.2 Volumes This analysis assumes a double guillotine pipe rupture in one of the four main steam lines upstream of the inboard MSIV and failure of all four main steam shutoff valves (1 N11-F0020A, B, C, and D) valves to close as a result of a common power failure (single-failure criterion).

The maximum allowable MSIV leakage of 250 scfh ls modeled to occur through two pathways:

(1) through the broken steam line and, (2) through the second and third intact steam lines. The volume of the ruptured main steam line between the MSIVs is 146 ft3 (DIN 11). Leakage past the second MSIV in this line is released directly to the environment.

The volume of the two intact steam lines between the reactor vessel and the inboard MSIVs is 440 ft3 (DIN 25). The leakage past the first MSIVs in these lines is released to the volume between the first and second MSIVs which Is 292 ft3, two times the volume between the MSIVs In one steam line (146 ft3) (DIN 11). Leakage past the second MSIVs In these lines is also released directly to

the environment.

This configuration was previously identified in DIN 17 to be limiting with respect to dose consequences.

8.2.3 Flows This calculation will apply a maximum MSIV leak rate of 250 scfh with the worst-case main steam line leaking no more than 100 scfh. The leakage limit is assumed to occur: (1) 100 scfh through the broken steam line, (2) 100 scfh through a second intact steam line, and (3) the remaining 50 scfh through a third intact steam line. As stated above, leakage is modeled to occur through two paths, one path consisting of the broken steam line and a second path consisting of the second and third Intact steam lines. All leakage past the outboard MSIVs is assumed to be released to the environment.

The drywell atmosphere will not be at standard conditions after the reactor blowdown.

The MSIV leakage rate must be.converted to a flow at the drywell conditions.

The MSIV leakage rates at the drywell conditions were determined from PSAT 04202H.04 Rev. 0 (DIN 12). The leakage rates are reduced at 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months when well-mixed conditions between the drywell and primary containment apply. Additionally, the flow In the main steam lines past the Inboard MSIV is represented as well-mixed.

The total MSIV leakage rates (DIN 1 and 11) are 298 cfh for the first two hours and 247 cfh thereafter.

The maximum flow rate is 191 cfh (DIN 11) through any single main steam line to the environment.

The values used in the model are given in Section 6.12. In addition to the leakage through the MSIVs, the drywell will also continue to leak activity Into the containment over this 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months period. This calculation will assume a leakage rate of 3000 cfm for the drywell bypass flow consistent with PSAT 08401T.03 (DIN 25). 8.2.4 Release Points All MSIV leakage past the outboard MSIV is assumed to be released directly to the environment.

No credit for holdup in the auxiliary building or turbine building is taken.

Page29 8.2.5 Model The RADTRAD model applied for this leakage path, as well as the containment airborne leakage path, Is illustrated in Figure 2. 8.3 Containment

& Containment Bypass Leakage 8.3.1

Volumes In addition to the main steam lines, the following volumes are used in the LOCA airborne leakage dose calculation (DIN 11 and DIN 31): Volume Name In In Model Model 1 Drvwell 5 Sprayed 6 Unsprayed 7 Annulus 9 Control Room Orywell Table 8-2 LOCA Volumes Description Sprayed Region of the Containment above the Operating Floor at El. 208'10n Unsprayed Reaion of the Containment Secondary Containment Control Room Volume (tr') 2.765E+05 4.812E+05 6.842E+05 1.96E+05 390,020 The volume of the Perry control room has recently been re-evaluated.

The current volume to be used in Control Room Dose calculations is 390,020 ft3 (DIN 16).

8.3.2 Flows From Drywell Volume Into Containment (Suppression Pool Bypass) The flow rate from the Drywell to the Wetwell are given below, see PSAT 04212H.02 (DIN 43): Time After Gap Release (hours) 0-0.5 0.5-2.0 2.0-720 TableS-3 Drywall Flows Flow from OW to WW (cfm) 0 3000 2.77E+05 Flow from WW to OW (cfm) 0 0 2.77E+05 Page 30 At two hours, the drywell and unsprayed portion of the containment will be assumed to become instantly well-mixed without credit for suppression pool scrubbing in accordance with Regulatory Guide 1.183, Section 3.7 (DIN 7). From Unsprayed and Sprayed Containment Volumes to the Environment By the time the gap release* begins, the containment is completely isolated and only containment leakage is assumed. The design basis containment leakage for Perry is 0.2% per day. Since the AEGTS will not completely draw down the annulus for 40 seconds, a 40 second pressure period is assumed in which all containment leakage is assumed to leak directly to the environment, but because there is no radionuclide release during this 40 second time period, before gap release which begins at two minutes, this leakage does not contribute to the onsite or offsite doses and is not included in the model. From Unsprayed and Sprayed Containment Volumes to the Annulus The majority (89.92%) of the total containment leakage (la) is drawn into the annulus by the AEGTS. Although the primary containment is enclosed by the secondary containment, there are systems that penetrate both the primary containment and the shield building boundaries that could create potential pathways through which fission products in the primary containment could bypass the leakage collection and filtration systems associated with the shield building.

The Perry Technical Specification SR 3.6.1.3.9 (DIN 15) limit the secondary containment bypass leakage to equal to or less than 5.04 percent of the primary containment leak rate. This analysis uses a bypass leakrate of 10.08 percent of the primary containment leak rate. . From Unsprayed and Sprayed Containment Volumes to the Annulus As stated above, the majority (89.92%) of the total containment leakage (la) is drawn into the annulus where it is filtered by the installed HEPA filters at a credited efficiency of 99% before being released Into the environment.

Page31 Mixing Between the Unsprayed Containment and Sprayed Containment The mixing rate between the unsprayed containment and the sprayed containment is 71,400 cfm, Calculation PSAT 04202U.03, Rev. 0(DIN13).

From Secondary Containment The only flow from secondary containment is via the AEGTS system which draws 2000 cfm through a charcoal-filter unit and HEPA filter unit. The HEPA filters are tested per Regulatory Guide 1.52 and therefore are credited for a 99% removal efficiency in the analysis; however, no credit is taken for the charcoal adsorbers in this analysis.

8.3.3 Removal Mechanisms Natural removal mechanisms for elemental iodine and aerosols will be applied in this calculation using NRC correlations.

Elemental iodine removal is credited in the drywell and containment volumes. Aerosol removal is credited only in the drywall and unsprayed region of the containment since containment spray will adversely impact the particle size distribution In the containment.

Fission product removal by containment sprays is considered.

The Perry containment spray system is initiated manually based on high radiation readings or is Initiated automatically approximately 10 minutes following a LOCA based on pressure and low water level. In this calculation, sprays are assumed to be manually Initiated at 30 minutes. The Powers model for aerosol removal by sprays which is built into the RADTRAD code Is used in this analysis.

Consistent with. the guidance in Section 3.3 of Appendix A to R.G. 1.183, the maximum spray decontamination factors for elemental iodine is 200 based on Standard Review Plan 6.5.2, Section Ill, 0. After the aerosol mass has been depleted by a factor of 50, the spray removal lambda is assumed to decrease by a factor of 10. The following section determines when these DFs were determined to occur. As discussed In Section 3.3 of Appendix A to R.G. 1.183, these OFs are based on the inventories at the end of the in-vessel release phase. Containment spray is assumed to end at 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and the aerosol removal by containment spray is terminated.

Decontamination Factor Reductions As discussed above, the elemental iodine removal by natural deposition is neglected after a OF of.200 is reached. Based on the elemental Iodine lambda of 0.878 hf 1 in the drywell, a OF of 200 would be reached in approximately six hours without any leakage or decay. The output in Attachment 2 indicates that the Orywell 2-hour post-accident release (I.e., 2.0000 hr) elemental 1-131 inventory of 2.4087E+22 atoms has reduced to 1.29E+20 at 3.0 hours0 days 0 hours 0 weeks 0 months post-accident representing a OF of 186. This calculation will therefore model the elemental Iodine removal to end at 3.0 hours0 days 0 hours 0 weeks 0 months in the drywell. In the containment, the total (sprayed+

unsprayed) elemental 1-131 inventory Is 3.1526E+21 atoms after the drywell is flushed at two hours post-accident (i.e., 2.0000 hr). The total activity in both regions of the Page 32 containment is considered because, if only the activity in the sprayed region of the containment Yias considered, a longer period of the higher lambdas would be applicable.

At 3. 75 hours8.680556e-4 days 0.0208 hours 1.240079e-4 weeks 2.85375e-5 months , the total elemental 1-131 Inventory in the sprayed and unsprayed containment regions is 1.6644E+19 atoms, representing a OF of 189. This calculation will model spray removal to end at 3. 75 hours8.680556e-4 days 0.0208 hours 1.240079e-4 weeks 2.85375e-5 months . The particulate removal (Powers Model) in the sprayed region of the containment is reduced by a factor of 10 when the aerosol activity is reduced by a OF of 50. In the containment, the total (sprayed + unsprayed) particulate inventory is 8.13 kg after the drywall is flushed at two hours. At 4.95 hours0.0011 days 0.0264 hours 1.570767e-4 weeks 3.61475e-5 months , the total aerosol inventory Is 1.64E-01 kg, representing a OF of 49.7. This calculation will model this removal coefficient to be reduced at 4.95 hours0.0011 days 0.0264 hours 1.570767e-4 weeks 3.61475e-5 months . This is accomplished t>Y reducing the spray flow used in the Powers Model by a factor of ten at this time. 8.3.4 Release Points All source terms released via containment leakage are released through the plant vent 8.3.5 Model The containment airborne model is illustrated In Figure 2 which is based on the time at which the gap release begins. This figure also includes the MSIV leakage transport pathways.

8.4 Control Room Although the current configuration of the control room HVAC system would automatically Initiate the control room recirculation on a LOCA signal, this analysis assumes that the CRERS is manually initiated at thirty minutes. Once the CRERS is initiated, CRERS fans recycle 27,000 cfm of control room atmosphere through HEPA filters and charcoal adsorbers before being returned to the control room. The normal control room recirculation air flow is 45,000 cfm (DIN 52) including 6,600 cfm of outside air for ventilation.

To represent the normal positive pressurization in the control room, the exfiltration air flow is modeled as 4,600 cfm before Isolation at 30 minutes. The RADTRAD code only allows a single control room recirculation air flow. As a result, the normal recirculation air flow Is not modeled. This is acceptable because the normal recirculation flow does not change the radionuclide concentration in the control room. After Isolation, unfiltered inleakage of 1375 cfm is assumed to be drawn from the control room intake duct for the duration of the postulated accident (30 days). The flow from the control room to the environment is also set at 1375 cfm to avoid pressurization.

Consistent with the requirements of R. G. 1.183, the contribution to the control room dose due to shine from the containment building (0.13 Rem) and release plume (0.002 Rem) must be considered.

These dose contributions are given in PSAT 04202H.13 (DIN 17). These 30-day doses to the control room were generated with the previous TIO 14844 source term that assumed an instantaneous release to the containment of 100% of the core Inventory of noble gases and 50% of the radioiodines.

These Page 33 assumptions are conseNative compared to the Alternative Source Term methodology (AST) due to removal of radlolodines from the containment atmosphere by sprays and deposition thereby reducing the radionuclide concentration in containment.

In addition, the total halogen release fraction is 0.3 for the

AST methodology providing additional margin. Based on these considerations, the previously calculated shine and plume doses are considered bounding for this analysis.

8.5 RADTRAD MODEL The models developed for the analysis are illustrated in Figures 1 and 2. 9.0 Operator Actions The operator actions assumed in this analysis include the following:

1. Manual initiation of containment spray at 30 minutes 2. Manual initiation of CRERS at 30 minutes 3. The pH calculation (DIN 26) assumes initiation of Standby Liquid Control (681:6) to control suppression pool pH SL r.. I&.> I. i1*1'1 SUPPRESSION POOL-1 114,379 ft 3 Figure 1 Fission Product Transport Model (ECCS Leakage Pathway) 1-Pool to Environment Page 34 0.03 2 s 24.5-end}

ENVIRONMENT -2 6. 71 cfm (24-24.5 hrs) 2 -Control Room Intake . 6000 cfm + 10% = 6600 cfm 1375 cfm after CRERS initiation CONTROL ROOM -3 390,020tt3

3 -Control Room Exhaust 4800cfm 1375 cfm after CRERS initiation Sprayed* 5 4.812e5 ft3 Figure 2 Fission Product Transport Model (MSIV and Containment Leakage Pathways)

Gap Release Phase 6

Sprayed ID Annulus .. 3-Unsprayed to Sprayef (4-Sprayed to Unsprll)'ed 11.400 c1m I Jl1.400 c:1m Annulus-7 1.96e5ft 3 9-AEGTS (AnruWs to EnvironmenQ

' 2000cfm 1-D!ywell to Unsprayed

3000 cfm Unsprayed.

6 6.842e5ft3 2* Unsprayed to Drywell after two hours I:

lo Annulus 8-Unsprayed lo Envfrmunanl . 7

Sprayed lo Environment Page 35 10-CR Intake 6000cfm+10%

= 6600 cfm 1375 cfm aftllr CRERS inlllallOn Environment.

a -. Conlrol Room

9 3S0,020ft3 DlyweD-1 2.765e5ft3 MSL2*3 MSL3*4 440 ftJ 1-----.i-292 ft 3 12* Drywall to 14-MSL.2 MSL2 L...-----1 to MSL3 MSL1 *2 146ft3 15

MSL1 lo Envlmrment 13-0rywell to L....----i 16-MSL.3 to Environment

11-CR Exhaust 4800cfm 1375 cfm after CRERS Initiation Page 36 10.0 COMPUTATION The RADTRAD output files, which include the Input summary, are given in Attachments 1 and 2. The* RADTRAD input and output files used for this calculation are identified below: Description ECCS Leakage Containment Leakage Plant scenario file Auxiliary RADTRAD Input Files Nuclide Inventory File Release Fraction and Timing File Dose Conversion Factors Output File PNPP ESF.psf PNPP ESF.nif pnpp_esf.rft Fgr11&12.lnp PNPP ESF .out Files for the TSC dose calculation are: Plant Scenario Flies PNPP LOCA.psf PNPP LOCA.nlf PNPP _DBA.rft Fgr11&12.inp PNPP LOCA.out Output Files PNPP ESF TSC.psf PNPP ESF TSC.out PNPP LOCA TSC.psf PNPP LOCA TSC.out For the TSC analyses, the nuclide inventory files, release fraction and timing files, and the dose conversion factor files for the LOCA and ESF cases are the same as above. 11.0 Overall Results Table 11-1 presents the dose results for individual leakage pathways for MSIV leakage, containment leakage, containment bypass, ECCS leakage, and shine dose. Control room shine dose is from DIN 17 and 25. Table 11-1 Dose Results (rem TEDE) Pathway EAB LPZ Control Room TSC Containment

& MSIV Leakaae 20.4 5.0 1.7 0.36 ECCS Leakage 0.79 1.83 1.15 0.05 Shine Dose 0.132 0.132* Total 21.2 6.9 3.0 0.5 Regulatory Limit 25 25 5 5 *Assumed to be the same as the Control Room CALCULATION ADDENDUM Page 1 0 BV1 I n BV2 l 0 DB I 181 PY TITLE/

SUBJECT:

(MUST MATCH ORIGINAL CALCULATION TITLE (SUBJECT))

Control Rod Drop Accident Radiological Analysis using Alternatllie Source Terms Classification 181 Tier 1 Calculation 181 Safety-Related/Augmented Quality D Nonsafety-Related Open Assumptions?

D Yes 181 No . If Yes, Enter Tracking Number Initiating Document (Perry Only) Referenced In Atlas? D Yes 181 No CPerry Onlv> Referenced In USAR Validation Database D Yes 181 No Computer Program(S)

Program Name Version I Revision Category Status Description NIA Originator (Print, Sign & bate) Reviewer/Design Verifier(Print, Sign & Date) s;g t:JJ2sl ,, A. Widmer.A.

6-24-14 Marvin Morris 6/2512014 jVi.

tf //;, L r t>. C.ull r I * µ . .,,D OBJECTIVE OR PURPOSE OFADDENDUM:

I The purpose of this addendum is to clarify assumption 3.1.5.8 with respect to the iodine fractions released, and iodine fractions input into the RADTRAD code. One objective is to prevent an error likely situation from occurring during the next revision.

SCOPE OF ADDENDUM:

There is a discrepancy*between Assumption 3.1.5,8 and the inplitinto the RADTRAD code. This addendum is initiated to ensure that the next full revision corrects the discrepancy and to formally document that there is no impact on the results or conclusions.

LIST NEW DOCUMENTS TO BE ADDED TO THE DOCUMENT INDEX (DIN); 8 c:i c z I!! -z .! ::J i5 Document Number/Title Revision, Edition, Date CD D. a: .E D D D D

SUMMARY

OF RE SUL TS/CONCLUSIONS OF .ADDENDUM:

D.: ; 0 D D There are no changes to the results of this calculation.

This addendum documents a discrepancy between assumption 3.1,5.8 and the input values into attachments 1, 2, and 3. The assumption states that the iodine release from the turbine and condenser is assumed to be 97% elemental and 3% organic and is consistent with guidance In section 3.6 of Appendix C to Regulatory Gulde 1.183 Rev. 0. Attachment 1 (pages 4, 9, 27, and 32) and Attachment 2 (pages 2, 9) used iodine fractions of 95% aerosol, 4.85% elemental, & 0.15% organic. Attachment 3 (pages 2; 8) used iodine fractions of 97% elemental and 3% organic. Attachment 3 utilized the correct release from the turbine and condenser as It matches the release fractions contained in assumption 3.1.5.8. Future revisions of this calculation shall assume the iodine fractions released from the turbine and condenser to be 97% elemental and 3% organic unless updated regulations dictate othelWise.

As stated above, this inconsistency does not impact the results or conclusions of Revision 1. For the dose associated with the EAB, LPZ, and Control Room (without Isolation), the form of the iodine released does not impact the dose to those areas as no filtration was credited in the calculation for the offsite locations or the Control Room (without isolation).

As a result, there Is no change in the consequences for those areas. For the Control Room evaluations which assume Isolation and filtration, there is also no impact on Control Room doses because the efficiencies for the Control Room filtration were set to 80% for both the charcoal and HEPA filters. This ensures that the elemental iodine, organic iodine, and aerosol iodine are filtered the same. There would be no change to the isolated control room dose unless the filtration efficiencies were changed such that the CALCULATION ADDENDUM Page2 removal efficiency for different iodine fonns are differeni.

As evidence, attachment 2 and attachment 3 of the base calcu!aiion utilized the different iodine fractions.

It is noted that the calculated EAB and LPZ doses are identical as seen oii Page 30 of Attachment 2 and Page 44 of Attachment LIMITATIONS OR RESTRICTIONS CREATED BY ADDENDUM:

None IMPACT OF ADDENDUM ON OUTPUT DOCUMENTS:

None DESCRIBE WHERE THE ADDENDUM WILL BE EVALUATED FOR 10CFR50.59 APPLICABILITY:

RAD/SCREEN/EVAL 13-02880 for Revision 1 of this remains applicable.

LIST SUPPORTING DOCUMENTS: (Include total number of pages) Design Verlfjcatlon Record 1 page Calculation Review Cttecklist 3 pages Design Interface Summary 7 pages LIST ATTACHMENTS: (Include total number of pages) None 3.2 CLASSIFICATION OF STRUCTURES, COMPONENTS AND SYSTEMS Certain structures, components and systems of the nuclear plant are considered important to safety because they perform safety actions required to avoid or mitigate the consequences of abnormal operational transients or accidents.

The purpose of this section is to classify structures, components and systems according to the importance of the safety function they perform. In addition, design requirements are placed upon such equipment to assure the proper performance of safety actions, when required.

3.2.1 SEISMIC CLASSIFICATION Plant structures, systems and components important to safety are designed to withstand the effects of a Safe Shutdown Earthquake (SSE) and remain functional if they are necessary to assure: a. The integrity of the reactor coolant pressure boundary, b. The capability to shut down the reactor and maintain it in a safe condition, or c. The capability to prevent or mitigate the consequences of accidents that could result in potential offsite exposures comparable to the guideline exposures of <10 CFR 100>. Plant structures, systems and components (including their foundations and supports) designed to designated as Structures, components, Class 1, Safety Class 2 or Safety discussion of safety classes) are except for (1) those noted in

a or <10 CFR 50.67> (future revisions to design basis analyses that compare consequences to 10 CFR 100 will be updated to <10 CFR 50.67>} 3.2-1 the event of an SSE are in

. designated as Safety e <Section 3.2.3> for a Category I (2) those portions of Revision 12 January, 2003 No changes to this page. Provided for information. !New Page ... I or B) is also deactivated and the electric heating coil in the charcoal filter train is automatically energized upon receipt of an emergency signal. The emergency recirculation system causes the supply air to be filtered through the charcoal filter train (M26-D001A, B) before being distributed to the control room. This system is idle during normal plant operation. During periods of loss of offsite power, emergency power will be supplied by the standby diesel generators. The degree to which the recommendations of <Regulatory Guide 1.52> are followed is given in

. The main components of this system are located in the control complex at Elevation 679'-6" and consist of two 100 percent capacity filter trains. Each filter train includes the following sequential components: demisters, roughing filters, electric heating coil, HEPA prefilters, charcoal filters, HEPA after filters, centrifugal fan, isolation damper, and check damper. The fans, filter elements and dampers are of standard industrial design, manufactured in accordance with Quality Assurance (QA) requirements of Safety Class 3, Seismic Category I items. The filter racks, frames and housing are specially designed to satisfy the system space requirements and also meet the above QA requirements. Design information for the major components in this system is listed in

. 6.4.2.3 Leak Tightness The control room system is designed so that, when operating in a normal mode (admitting outside air), the system automatically maintains a positive differential pressure between the control room and the outside 6.4-9 Revision 12 January, 2003 The design basis radiological calculations for a assume an* unfiltered lnleakage of fmfor 30 minutes, which shows Normally, the control room boundary inleakage is maintained at a value consistent with pre-operational testing such that the actual inleakage is substantially less than 1375 cfm. Throughout the life of the plant, various plant activities may need to be performed which temporarily degrade the control room boundary such that the unfiltered inleakage significantly exceeds 1375 cfm. postulated LOCA uere te eeeyr YReer taese eaReitiaRs, parametrie that it is acceptable to delay the restoration of room boundary, provided that once it is restored, the actual inleakage would be below 1375 cfm for the remainder accident. to occur without This allows degradations of the boundary accident dose to the control room operators. to at or 6.4.4.2 are utilized during planned boundary can <l!'iE]IUQ 9 I 4 the bounding 4 ( > aae EJIU9 9 I 4 4 ( 2 ) > I lrapldly I to No toxic materials which could interfere with control room occupancy are stored in the plant. Sodium hypo-chlorite, rather than chlorine, is used as a biocide. No chlorine is stored on site. The potential effects of offsite and onsite hazardous materials are discussed in <Section 2.2.2> and <Section 2.2.3>. Protection against offsite toxic gases are detailed in <Section 6.4.1.g>. 6.4.4.3 Control Room Emergency Recirculation System The general arrangement and control of the control room emergency recirculation system is as described in <Section 6.4.2.2.2>. Detailed information concerning the emergency filter is presented in <Section 6.5.1>. The equipment is shielded, housed in a Seismic Category I structure, separated, redundant, and powered from the 6.4-15 Revision 14 October, 2005 the design basis LOCA <Section 15.6.5.5.1.9> credits an 80 percent removal efficiency of elemental and organic iodines by the charcoal filters in the CRERS. The Steam System Piping Break Outside Containment <Section 15.6.4>, Control Rod Drop Accident <Section 15.4.9>, and the Fuel Handling Accident <Section 15.7.4> and <Section 15.7.6>, do not take credit for the charcoal filters in the CRERS. organic species of iodine. For the CRERS, eerHI l.QCA aRalysis aREl eRe fyel RaREl:liR§ aeeiel:'eR6 seRsieiviey ease assYH1eel: aR eleHleRtal aREI: er§aRie reH1eval effieieRey ef eRly 90% fer 6Re sRareeal ael:sersers. For the FHAES, the alternative source term FHA analysis took no credit for the charcoal adsorbers. The CRERS and FHAES charcoal adsorber beds are 2 inches deep. Exhaust air for both plenums is maintained at less than 70 percent relative humidity. The HEPA filter efficiency of all the plenums is 99.97 percent on particles 0.3 microns and larger. Has eal'teR fer fyel RaREl:liR§ aeeiel:eRe. Additional bases for the design of the C RS, FHAES and AEGTS are presented in <Section 6.4>, <Section respectively. 6.5.1.2 System* Design 9.4.2>, and <Section 6.5.3> the LOCA analysis only credits the HEPA filters in the AEGTS and CRERS at an efficiency of 99 percent. The other design basis radiological calculations do not take credit for the t:tEPA filters In the AEGTS, CRERS, or FHAES. The design features of the CRERS, FHAES and AEGTS are compared to the recommendations of <Regulatory Guide 1.52> in

,

, and

respectively. Design of the activated charcoal adsorber plenums used in the CRERS, FHAES and AEGTS follows the guidelines of <Regulatory Guide 1.52> and ERDA 76-21. Each charcoal adsorber plenum contains the following:

a. Demisters to remove large particles and water droplets (about 1 micron diameter) . b. Roughing filters to remove large particles (about 1 micron). 6.5-4 Revision 13 December, 2003 No changes to this page. Provided for information.

!New Page ... I e. Gasketless activated charcoal adsorber beds to remove gaseous elemental and organic iodines. f. HEPA filters downstream of the charcoal beds to remove charcoal particles that may be entrained in the air stream. g. A fan external to the plenum. h. Instrumentation.

i. Test ports. j. Water deluge system for fire protection.

Plenum housings and filter support frames are shop fabricated. Potential leakage and bypass paths are closed by seal welding. No caulking or sealant is used. Housings are fabricated of carbon steel sheet. Filter support frames are of unpainted stainless steel. Roughing and HEPA filters are mounted in frames in accordance with the recommendations of ERDA 76-21. The activated charcoal adsorber is bulk loaded into the permanently installed, gasketless adsorber section which is seal welded to the housing and support frames of the plenum. Tray type activated charcoal adsorber units are not used. Spent charcoal adsorber material is vacuumed from the bottom or top of the plenum and is loaded into 55 gallon drums for shipment off site. New charcoal adsorber material is added at the top of the adsorber section. Personnel are not directly exposed to potentially contaminated adsorber material during the changing procedure. 6.5-5 Revision 12 January, 2003 6.5.1.6 Materials No changes to this page. Provided for context. jNew Page ... I Estimated quantities of materials used in the activated charcoal adsorber plenums for CRERS, FHAES and AEGTS are listed in

,

, and

respectively. The governing specifications for the various materials are also listed and provide information regarding chemical composition of materials used. There are no radiolytic or pyrolytic decomposition products from the ESF filter systems. Actuation of the activated charcoal adsorber plenum water deluge fire protection systems will extinguish a charcoal fire before pyrolytic decomposition products are formed. None of these systems are located in areas where gamma radiation sources are sufficiently strong to cause radiolytic decomposition products. Therefore, decomposition products do not affect any engineered safety features. 6.5.2 CONTAINMENT SPRAY SYSTEM 6.5.2.1 Design Bases a. The containment spray system (CSS) is a part of the residual heat removal (RHR) system.

b. The CSS provides containment cooling following a loss-of-coolant accident, in addition to being a fission product removal mechanism.

Refer to <Section 6.2.2> for the heat removal function of the CSS. c. The CSS consists of two completely redundant and independent loops. d. The CSS is designed to remain operable in the containment accident environment, which is discussed in <Section 3.11>. 6.5-9 Revision 12 January, 2003 INew Page ... I are not subject to clogging by particles less than 1/4 inch in maximum dimension. Each nozzle header is independently oriented to ensure efficient coverage of the containment volume. The minimum water supply flow rate to the containment spray system is 5,250 gpm. There are no spray additives for the CSS (other than the pH buffering ' chemical, boron solution, from the standby liquid control system, which is injected into the reactor vessel and suppression pool following a design basis LOCA). The CSS will automatically initiate after 10 minutes of a LOCA signal if containment pressure exceeds the high pressure setpoint. If containment pressure is less than high pressure setpoint, the control room operator can actuate the system manually. The sprayed and unsprayed volumes and regions of the containment, with their associated mixing rates, are discussed in <Section 15.6.5>. The CSS takes no credit for ventilation. 6.5.2.3 Design Evaluation No changes to this page. Provided for context. The containment spray mode of the RHR system is safety-related and is designed to operate following the postulated design basis loss-of-coolant accident. A high degree of system reliability is maintained through system quality control, by general equipment arrangement to provide access for inspection and maintenance and by periodic testing. A single failure analysis of the RHR system is given in <Section 6.2.2>. Because of the large volume of the containment atmosphere swept by the sprays, the spray mode serves as a removal mechanism for fission products postulated to be dispersed in the containment atmosphere following an accident. Radioiodine in its various forms is the fission 6. 5-11 Revision 12 January, 2003 product of primary concern in the SDl) ---.. accident. The major benefit of the tainment spray is its capacity to collect the containment atmosphere thus to the environment. Offsite and control room operator doses are a function of both the rate of removal and the final equilibrillm decontamination factor. The dose calculation assumes (non-mechanistically) that the containment spray will operate for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . However, the dose calculations also expand on this assumpti9n, noting the following: 11 The dose calculations assume the sprays are run for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , then are suspended. This is the most important time period for scrubbing of radiation down into the suppression However, in an actual event, spray use would not necessarily be suspended at 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , if appropriate conditions for their use still existed. Therefore, the phrase "up tou_is !!2!:_ intended to be interpreted to stop using sprays after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . 21 The phrase "up tou .!! intended to mean that in an actual event, the sprays will be run when it is appropriate, and not necessarily the entire time during the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of a LOCA. This does not invalidate the assumptions in the dose calculations. The accident guidance to operators must be written to be symptom based, rather than event based. Most postulated LOCAs will not result in .large radiation releases. Therefore, it would not be appropriate to run containment sprays for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following such an event. Another critical factor in spray use is containment pressu_re. Use of the sprays will work to reduce containment pressures, due to steam condensation and the containment heat removal function that they provide. In the majority of cases, if a high radiation signal is present from the containment radiation monitor and pressures are elevated in containment, the sprays would be run. However, if containment pressure gets reduced to near zero and use of the sprays is terminated by the operators, this does not have an 6.5-12 Revision 12 January, 2003 M.08..a-\ rM tic. u..' °""' e.. (1a.e.ro . 's bu.\ l .\-\"To Elemental Iodine removal is credited in the drywall and containment volumes. Airborne elemental iodine is removed by deposition to the walls In the drywall and containment. As reported in Section 5.1.2 of NUREG/CR-0009 (Reference 1 ), this process is driven by the temperature differences between the surfaces and the atmosphere. 'mpact on offsite doses (or the dose calculations) since pressure for containment and MSIV leakage has been The dose calcs assume that the.maximum allowable corresponding to the peak postaccident pressure (Pa) s during the entire 24 hourJperiod, so if containment re actually gets reduced to substantially less than Pa, a tion in leakage and the resultant offsite doses will follow. 1 Iodine Removal Performance Evaluation e,\e.M.tA'\-o.\) or!:;'&.'l\.l c-/ N\.l. pw (l>.e.r"(.tSO \) analysis is based on the assumptions presented below

. lM.A. u.t\Sf nalysis uses the flow associated with only on RHR pump operating in containment spray mode. It is conservatively ssumed that the system directly sprays approxima ely 41 percent of the free volume ovides a description of the The calculated iodlne removal in

fe;r t;l:le elemeRtal aREi pu*tie'c:llate iea!Res as uell as etl:le;r pa;rtie'c:llates. QeeaYse ef t;ae S'c:lFfaee a;rea ef tae iRitially aiiE'SEliE'RQ pa;rtie'c:llate, t;ae elemeRtal ieEiiRe is asswaea t;e se aase;rsea QRt;Q tRe pa!'tie'c:llat;e aREi te SQ !'em9'}QQ 11it;R it;, It has been conservatively assumed in these effectiveness that organic iodine forms are Evaluation of Analytical Assumptions 6.5.2.3.2.1 Iodine Retention by Spray Solution spray removal The equilibrium between the concentrations of iodine in the liquid are vapor phases is given by the partition coefficient, H, which is a AAl>TAAb "'-* --------c.oA.e," r'*' bore.. Revision 12 ft\ .(..'\i>c'o\t.,... lo,S-'\">. 6.5-13 January, 2003 c..ve.*l=.C.\ \:i '-\O ct..f+.u-b.u,."-".e.-p\ e...--W. fuc..{or c:{ S-o, f"'".\!\ \:; -6 ,,.+ 1...'\ \\.cl(,1"$ .... function of iodine concentration, pH and temperature. In accordance with (Reference

3) re-evolution of iodine does not have to be considered, (i.e., H will be very large) as long as the pH of the suppression pool is maintained greater than or equal to 7.0

postaccident.

6.5.2.3.2.2 Elemental Iodine and Particulate Removal Constant The ealsYlatieaal meelel Ysea te elete:l'llliRe the eleRl9atal aael paEtieylate ieeliae EemeTJal eeastaat is PelestaE Afplieel Teehaele§y's "STARNAUA" ee111pYter eeae (RefeEease 4} uhisa iaseEperates tae spFay remeTJal meaeliR§ feat;yi;es §iTJBR iR AppeREliH E ef (RefB:li9R99 d) I Tae iRpYt aata ey tae se111p1:1tei; eeae ie §iTJeR iR <Tasle ' 5 9>, Tae meaR El:liep fall aei1Jat ef feet *,1ae saleYlatee ey takiRIJ a uei§Atea a*JeEa§e ef the aei§ht ef eash EiR§ aseve tae eperatiRIJ fleei; aae the assesiatee sp£ay fleH Fate as £elle*,1e: i..-Re£e1 "" 45.15' """ 94.15' """ aRel m FiRIJ flew Eate a'l:HRSeE' ef aeaales u fleu Fate peF aeeele """ """ """ Tae spatial aaa tempeFal elistEi91:1tieRs are eleFiTJee fFem aaalysis YsiR§ tae a sempYteE' eeele "GPIRT" (RefeFease l) , 6.5-14 Revision 12 January, 2003 Section 6.5.2.3.2 2 The Control Room-Emergency Recirculation System (CRERS) subsyste each ha a high efficiency particulate air filter, charcoal adsorbers, and HEPA filter. he CRERS is an ESF system that Is tested in accorda with R.G. 1.52 (Re ence 4). The calculation model (Reference

6) as med an elemental and or le Iodine removal efficiency of 80 percent the charcoal adsorber removal e lency. Each HEPA filter is taste show a penetration an stem bypass of less than 0.05 percent when tested in cordance with Re atory Gulde 1".52 (Reference 4 ). A penetration a bypass of I than 0.05 percent allows credit for a particulate removal efficienc f 99 p ent per Regulatory Guide 1.52. The analysis therefore used a CRERS lter efficiency of 99 percent for aerosol particulates.

The AEGTS Includes HEPA fll rs and 4-lnc eep charcoal filters. Particulate removal by the HEPA filter s assumed to be 9 ercent in accordance with Regulatory Guide 1.52. e analysis conservative assumed a removal efficiency of 0 percen or the charcoal adsorbers. A simplified mo for estimating the fission product aeros emoval by containment rays following a postulated LOCA was used In e analysis. The model for rosol removal by sprays built Into the RADTRAD co is the Powers model. he Powers model was derived by correlating the results Monte Carlo unc ainty sampling analyses assessing the uncertainties In aeroso operties, a sol behavior, spray droplet behavior, and the initial and boundary c dltlons xpected to be associated with a postulated LOCA In the containment. '" 6..\oo'1ca.-- \.s. lV\. (,!'('" U..S.JT(L b.i ptV'\-c..f *"'-'s -h.r \t b..c:.+; '-V\.d... -k. &.lao'llL-wa.s \"'a. -h>r S'. 2., 'b.i,,S.jt'\. t 6.5.4 ICE CONDENSER AS A FISSION PRODUCT CLEANUP SYSTEM This section is not applicable to PNPP. 6.

5.5 REFERENCES

FOR SECTION 6.5 1. Postma, A. K.; Sherry, R. R.; Tarn, P. S.; "Technological Bases for Models of Spray Washout of Airborne *Contaminants in Containment Vessels," <NUREG/CR-0009>, October 1978. 2. ANSI/ANS-56.3-1979, "American National Standard for PWR and BWR Containment Spray System Design Criteria." 3. Electrical Power Research Institute, "Generic Framework for Application of Revised Accident Source Terms to Operating Plants," TR-105909, Interim Report, November, 1995. 4.... PeJ:estaF Applieel TeeR.ReleEjy, !Re,, "STAAWAY:A, A Geele fe;r; EvalYatiR(j . Se 1 J'eEe AeeieleRt Ae;r;esel 8eeav=ie;r; iR WaeleaF Peue: PJ:aat GeRtaiRR\8Rts1 A Geae QessriptieR aaa ValiaatieR aRa VeFifieakieR

4. Regulatory Guide 1.52, "Design, Inspection, and Testing Criteria for Air Fiitration and Adsorption Units of Post-Accident Engineered-Safety-Feature Atmosphere Cleanup Systems In LlghtWater-Cooled Nuclear Power Plants",.

Revision 2, March 1978. 5. NUREG-0800, "Standard Review Plan (SRP) 6.5.2, Containment Spray As A Fission Product Cleanup System," Revision 4, March 2007. 6. Calculation 3.2.15.16, "Design Basis LOCA Dose Evaluation Using Alternate Source Terms," Revision O Oeteber 20ts. 6.5-24 Revision 12 January, 2003 TABLE 6.5-9 INPUT PARAMETERS FOR THE SPRAY REMOVAL ANALYSIS Unsprayed containment volume, F 3 Mean spray fall height, ft Number of spray pumps operating Spray flow rate, gpm Spray solution pH Q, Spray Flux, cfm/ft2 Alpha, unsprayed/sprayed volume Pct, uncertainty percentile Geemetrie meaR Elrep siae fer spatial ElietrisYtiea 1 em Geemetrie meaR partiele eiae fer iaeemiR§ ae;eesel 1 em Geemet;eie meaR staRElare eeviatieR Ne uall eeR9easatieR Ne eeaEieRsatieR eR HateE Elreplete Ne eeaE;:i.Ele;eatieR ef pailtiele

&ly§reseepieity 6.5-56 481174 684226 1 *5, 250 7.0 0.0621 1.422 10 Revision 12 January, 2003 Isotope I-131 I-132 I-133 I-134 I-135 Kr-83m Kr-85 Kr-85m Kr-87 Kr-BB Kr-89 Xe-13lm Xe-133m Xe-133 Xe-135m Xe-135 Xe-137 Xe-138 Time Period (hr) 0-8 8-24 24-720 NOTES: CRDA, MSLB TABLE 15.0-4. DOSE CONVERSION FACTORS 111 Thyroid (rem/Ci) 1. 49E+6 5.35E+4 3.97E+5 2.54E+4 1.24E+5 Breathing Rates Whole Body 0.25xMeV/dis

8. 72E-2 5.13E-1 1.55E-1 5.32E-l 4.21E-1 5.02E-6 3. 72E-2 5.25E-4 1. 87E-l 4.64E-1 5.25E-1 2.92E-3 8.00E-3 9.33E-3 9.92E-2 5.72E-2 4.53E-2 2.BlE-1 111 The lowing dose conversion factors (DCF's) are used in the alternative term analyses; -QCF's fer iAAalatiea:

1989 (Reference

-CEDE: EPA Federal Guidance Report 11 -1989 (Reference

11) DOE/EDE: MACCS2 computer code (Reference 12), which used Federal Guidance Report 12 -1993 (Reference 13). 121 This breathing rate was used for the duration of the Control Room radiological consequence analyses.

15.0-37 Revision 13 December, 2003 I. . TABLE 15.4-12 CONTROL ROD DROP ACCIDENT EVALUATION PAR.1\M&TERS Data and assumptions used to estimate radioactive source from postulated accidents.

A. B. c. D. E. F. Power level Burn up Fuel damaged Release of activity by nuclide Iodine fractions, % (1) Organic ( 2) Elemental (3) Particulate Reactor coolant activity before the accident.

Scenario 1 Assumptions 3,833 MWt N/A rods.ut

Scenario 2 Assumptions 1376 N/A 4.85 95 N/A N/A II. Data and assumptions used to estimate activity released. A. B. c. D. E. F. G. Condenser leak rate (%/day) 1.0 Turbine building leak rate (%/day) N/A Valve closure time (sec) N/A Adsorption and filtration

efficiencies (1) Organic iodine N/A (2) Elemental iodine N/A (3) Particulate iodine N/A (4) Particulate fission products N/A Recirculation system parameters (1) Flow rate N/A (2) Mixing efficiency N/A (3) Filter efficiency N/A Containment spray parameters (flow rate, drop size, etc.) N/A Containment volumes N/A 15.4-44 N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A Revision 12 January, 2003 The RAST _analysis

'e 1 3 9 p 11 rs 11 ed initially t 0 s 11 pp 0 rt an jncrease in the main steam line leak rate to 250 scfh and to eliminate the MSIV leakage Control System The is based on the following:

0 *

main steam lines, Regulatory Gulde 1.183

controlling the pH of the water in the containment to prevent iodine re-evolution,

operating the containment spray system for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months <Section 6.5.2.3>,

not crediting iodine removal by charcoal adsorbers in the Annulus Exhaust Gas Treatment System (AEGTS),

delaying actuation of the control room emergency recirculation for up to 30 minutes, elemental and organic iodine removal efficiencies control room emergency recirculation system charcoal adsorbers 9 5 percent to 50 percent, .

increasing the engineered safety feature system leakage outside The SA.S.:r. analysis considers the following four potential fission product release pathways following the design basis LOCA:

main steam isolation valve leakage,

containment leakage, 15.6-23 Revision 12 January, 2003 The analysis conservatively assumes that the fission product leakage from the main steam lines is released directly into the environment.

The leakage past the MSIVs is conservatively assumed to begin immediately after the accident. In actuality, the three intact steam lines would contain trapped steam which would be relatively cooler and more dense a*s compared to the atmosphere in the reactor vessel upper head during the overheating of the core. This condition would greatly inhibit mixing between the activity released from the core and the steam leaking through the three intact steam lines and the three*associated sets of MSIVs. However, for conservatism, all of the lines are assumed to be leaking contaminated drywell atmosphere. Other significant conservatisms in the analysis of steam line transport include: S't"E.'T" ... 0 consd<eratiat pf lipe mass leak rate with (1) t 'I\ -&, ro,,:J D"

tM.,..

C'\$l 'Is (no Ct'\. -flow (2) of No consideration of particulate removal and even plugging of the extremely small MSIV leak paths due to particulate deposition at the entrance to or within the leak path as the gas flow Two configurations were analyzed to cover all single-failure possibilities. In the first configuration (Configuration 1), the inboard MSIV on the affected line was assumed to fail open, and this line was assumed to leak at 100 scfh. The three intact lines were then assumed to leak at 100 scfh, 50 scfh, and 0 scfh to maximize flow rates through the lines, which in turn maximizes the activity release. At 20 minutes after the start of release the third safety-related and seismically-qualified isolation valves (just outboard of the outboard 15.6-25 Revision 12 January, 2003 (V\A.l{\ tr N&-s.) > j \ Elemental iodine retention efficiency is based on resuspension rates from (Reference the main steam lines to retain aerosol fission products) was slightly reduced in the analysis. This aerosol removal efficiency is equivalent to an increase in aerosol penetration of 10 percent. This was done to further increase steam line pathway. , 5h"A.\ \v -ti> r(l...&..w..-\loV\.J -t\.-e...

15. 6. 5. 5 .1. 2 Fission Product Transport in Drywell t oA1 V\.Q.-o ('Q..---\e.t'\.""OI\

WC'-' rQ.bcacl_ The most limiting OBA, with respect to the offsite and control room \O°b .for radiological consequences, is considered a large-break LOCA of a double guillotine pipe rupture in one of the four main steam lines "\-D** '\ upstream of the inboard MSIV. It is further conservatively assumed that all fission products are released directly to the drywell and leaked into the primary containment and into the main steam lines, bypassing the suppression pool. The analysis also assumes that at a point two hours after accident initiation (when the ECCS is assumed to be able to reach the core and reflood it) the fission products are homogeneously distributed between the drywell and the primary containment. The objective of this well mixed approach is to achieve an appropriate balance for the design of drywell leakage mitigative devices such as the MSIVs as well as containment leakage mitigative features such as annulus exhaust gas treatment system. Reference 19 HEPA filters In the in-vessel fission product releases terminate 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after accident initiation. For the fission product releases to terminate, the reactor vessel would need to be reflooded. In lieu of evaluating all of the potential steaming rates due to various reflooding scenarios, the analysis assumes that a substantial amount of fission products will end up in the primary containment as well as in the drywell, and as such, 15.6-27 Revision 12 January, 2003 1exr*t 1 mltigati ve features such as the HEPA filters in the annulus effl 11ent gas treatment system are designed to acconunodate a significant portion of the source term. The 2-hour assumption for the homogeneous mixture of the source term between the drywell and the containment is used since it provides an appropriate balance, because the "worst 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months " are considered for the EAB radiological dose results, as opposed to simply the first 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months as was done when the TIO source term was used. The radiological consequences are dependent upon the drywell bypass leakage prior to the termination of fission product release at 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . Because of this sensitivity, the analysis uses a steaming rate of an intact core without relocation to the lower head region, on the order of 3,000 cfm. For the period prior to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , the analysis conservatively does not credit steaming due to relocation, cooling from alternative water sources, or the release of hydrogen gas, all of which would provide a higher steaming rate and remove more of the fission products from the drywell region. Aeroso Deposition r.I.:itbi n B!!}Well Cl Activity released to the drywell as a result of the design basis loss-of-coolant-accident ls initially airborne and can be removed from the atmosphere in one of four ways: (1) Convection from the drywell to the containment (2) Natural removal within the drywell (e.g., particulate sedimentation) (3) Leakage into the broken steam line and through the MSIVs (4) Leakage back into the reactor vessel and through the MSIVs 15.6-28 Revision 12 January, 2003 Elemental Iodine moval Is credlte In the drywall volu lrbome element.al lodloe la removed by deposition to the waflso . This process Is driven by the tempe re differences elween the surfaces and the atmosphere. The cafcutated retoval constants are applied untll a decontamlnaUon faotori {OF) of 00 has been obtained. Aerosol removal

In th rywel s modeled using the Powe,as removal model as given In l:t-6189 (Reference 20). The lower * . bo <fecon amlnatlon coemctent associated with the 1oth ercentlle uncertainty Wb used for conservatism.

The leakage cont rt hut! on is by .design/ and therefor pri cipal for depletio of activity in the drywall tmosphere (ot er than by radioactive decay) is convection from the drywall . . con ainment and natural removal w the drywall. de letlon due to MSIV Following the fuel releas'! phase accident, the restoration of ECCS (thus arresting further core damage) would quench the core debris, and results in a rapid sweep-out f the drywall into the containment as discussed in Section 5.2.3 of (Re erence 18). For the design basis analysis, a egotiated licensing basis was established for the transport of cti.vity between the containment and the drywall. The negotiated basi in effect mixes activity between the regions and does not consider a s eep-out of the activity after two hours. The negotiated parame

. Na*tnrel remaual gf ectisrlty dnp to phy 0 sjcat pCOCPP' (i p 0 gthpr than hy, radioecttve decay) can hp pssnct at9d wt th mepy pffect*,, inc1J1d1 ng processes (described 1p Sect'n" S ? 3 and nppppdty r gt (Reference

18) are credtted in this apa 1 yeis The Po 1 estar Dpp 1 ierl Techgnlogy "STPRNDllP" comp 11 ter code (Reference JS) ts used for +be ce>cnlftt1o 0 of 15.6-29 The kear 1 np 11 t pssnmpttnn" Revision 12 January, 2003 ..

b'I o+ ' °'-k.s £:* I.. T'r,;,:'" 2. h.out93 Eates aEe alee assameel te afilply t;e elemeRtal ieaiRe {see SeetieR ef {RefeEeRee lll)), )1ete taat tae 15TAR):IAYA aRalysls e9Rsieen fleu eat gf tae 9Eyuell aaa seaimeat:atieR si111Y:ltaReeasly, IR this uay the Eelll9val Eates (whieh impEe'.'e uith iRe£easiRIJ fi1a£tiealate eeeeeet£al;ieR) aEe R9t 9V9EestimateEI:, The pa£tiealate Eelease fEeRI the eEy11ell (assgeiateEI: uith the elEyueH te eeataiRmeRt eee*:eetieR eliseassea ase\*e) seeemes the iRpat feE the STAR):IAYA ealealat;ieR feE the spEayeel Eel]i9R ef the 80RtaiRmeRt <SeetieR afteE BBiRq EBQ1:188Q sy a faeteE ef a u 4 4 l;e a8891:1Rt feE the faet t;lut the Bfi1£aye9 liB§i9R ie eRly 41% Bf the eeataiameat f£ee vel1:1me, HeEe a§aiR, the iateat is te eRB1:1£e that the pa£tiealat;e eeReeRtFatiea (aael theFefeEe, the Fate ef paEtiealate Eemeval) iR the spEayea Eeljiea ef the eeRtaiRmeRt is Ret e\*eFestiRlatea. 15.6.5.5.1.4 Containment Leakage Pathway The primary containment consists of a drywell, a wetwell, and supporting systems to limit fission product leakage during and following the postulated LOCA with isolation of the containment boundary penetrations. The design basis leak rate of the primary containment is 0.2 volume percent per day. The analysis (30 days). for the at 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months as permitted by <Regulatory Guide 1.183> the design basis reduces remaining The.secondary containment (shield building) which surrounds the primary containment will collect and retain fission product leakage from the primary containment and will release fission products to the environment in a controlled manner through the AEGTS. AEGTS will maintain the secondary containment pressure negative following a OBA by the time the gap release could migrate outside the containment structure. Therefore, if a short period of time exists post-LOCA when the annulus pressure is not negative, the dose calculations would not be affected. Although the primary containment is enclosed by the secondary containment, there are systems that penetrate both the primary 15.6-30 Revision 12 January, 2003 containment and the shield building boundaries that could create potential pathways through which fission products in the primary containment could bypass the leakage collection and filtration systems associated with the shield building. The analysis conservativeiy assumes 10.08% of the primary containment leakage bypasses the secondary containment (the Technical Specifications limit bypass leakage to a lower limit). The analysis assumes 99.92 percent of the primary containment leak rate goes into the secondary containment for its radiological consequence analysis. This leakage is collected in the shield building and processed through the AEGTS HEPA filters before being released into the environment. The remaining 10.08 percent of the primary containment leak rate is assumed to bypass the shield building and to be released directly to the environment for the entire duration of the postulated LOCA. 15.6.5.5.1.5 Annulus Exhaust Gas Treatment System The AEGTS is an engineered safety features system and is designed to collect, process, and release the fission product leakage from the primary containment into the shield building. The AEGTS is a redundant system consisting of two 100 percent capacity subsystems. Each subsystem has a design capacity of 2000 cfm and consists of, among other things, a HEPA pre-filter, one 4-inch deep charcoal adsorber, and a HEPA post-filter. The system is designed to Seismic Category I standards and is located in a Seismic Category I structure. The system is operated continuously during normal plant operation, and . it maintains a slight negative pressure in the shield building. The analysis assumes a 99 percent removal e iciency for fission products in aerosol form for HEPA filters. The analysis however does not consider any by the charcoal adsorbers in the AEGTS.1\...t.. l\t\o..ll.til. So '2c.O<> c{!."" .flci"o l

\-o wN*1'. f\6

("'6\l ) o4 5.6-31

15. 6. 5. 5 .1. 9 Control Room Habitability Upon receipt of an ESF actuation system signal or high radiation, the control room Heating, Ventilation, and Air Conditioning (HVAC) system is designed to automatically switch to the emergency recirculation mode of operation (CRERS). The analysis conservatively assumes a 30-minute delay in actuation of the CRERS. The CRERS is a redundant system and each subsystem has a design flow capacity of 30,000 cfm. The analysis uses a conservative recirculation flow rate of 27,000 cfm. Each subsystem consists of, among other things, a High-Efficiency Particulate Air (HEPA) filter, charcoal adsorbers, and a HEPA post-filter.

The analysis also uses a conservative HEPA filter efficiency of .Q& percent for aerosol particulate charcoal ff removal efficiency for iodine in element land organic forms. an 80 99 ---------During normal operation, the HVAC system is designed to p essurize the control room envelope with 45,000 cfm recirculation airfl wand with 6,000 cfm outside makeup air. During an emergency, when he system operates in the emergency recirculation mode, the outside makeup air is isolated and the control room envelope is not pressurized relative to adjacent areas. To be conservative, the analysis uses cfm I inleakage to the control room during the emergency recirculation mode fe£ the eati£e 9Y£atieR ef the aeeieeat. The major parameters and assumptions used in the analysis are lis ed in

. first 30 minutes, followed by 1,375 cfm unfiltered lnleakage In the aeeeptaRee e£iteEiea ef <ilQ Gli'R 9Q 1 AppeaEliu A>1 GeaeEal Qesi§R G:dteEia 1Q 1 "GeAt£el The exemptieR peERlits w.se ef a 9 ESHI TEQE aeeeptaaee eEiteEia iR lie:Y ef "a £em 1:hele seEly, eE its BEIYh*aleRt te aRy pa£t ef the eeEly," as eY££eatly stateEl iR CQG 19 fe£ the eeRtEel 15.6-35 Revision 12 January, 2003 15.6.7 No changes to this page. Provided for context. REFERENCES FOR SECTION 15.6 INew Page ... I 1. Moody, F. J., "Maximum Two-Phase Vessel Slowdown From Pipes," ASME Paper Number 65-WA/HT-1, March 15, 1965. 15.6-43a Revision 14 October, 2005 TABLE 15.6-12a LOSS-OF-COOLANT ACCIDENT PARAMETERS AND ASSUMPTIONS USED IN RADIOLOGICAL CONSEQUENCE CALCULATIONS MAIN STEAM ISOLATION VALVE LEAKAGE PATHWAY Parometor Reactor power f"t04:, )(I ,0.2.) .3.15.B. MWt Drywell volume 2.765 K 10 5 t 3 6 volume 1.165 x 1 t Volume of one main steam line between MSIV's 146 ft 3 ]lnlnmetr1c flow rate, drywel 1 to al 1 m3jn steam ljnes (total leakage) JlQJ umptri C fl Clef rate (m3yimnm) d QDp main steam 1 i ne t*n pny1 roumpgt 29Q cfb frrng t -a to t e 7484 seconds 247 cfh frrnn t = 1484 seconds to 30 days 191 cfm .__---iVolumetrlc flow rate, drywall to broken steam line Oto 7484 seconds 1.987 ft 3/mln 7484 seconds to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 1.647 ft 3/min 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to 30 days 1.371 ft3/mln Volumetric flow rate, drywall to Intact steam lines Oto 7484 seconds 2.98 ft3/mln 7484 seconds to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 2.47 ft 3/mln 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to 30 days 2.056 ft3/min V 111 \ c.. f ra.t .e.-( l"l it'\ 3 .. l 3 { f:?'j rr.. i I\ Dt\Q... \CV\.4..1 \ . l'l\S \V .s J -io '\: ..: o to ,30 fi,..\-(!.... M.bd" &.\, 115 iY\S.\"'.s) t "" 0 tu 30 S 15.6-58 Revision 12 January, 2003 TABLE 15.6-12b LOSS-OF-COOLANT ACCIDENT PARAMETERS AND ASSUMPTIONS USED IN RADIOLOGICAL CONSEQUENCE CALCULATIONS CONTAINMENT LEAKAGE PATHWAY 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months -3 days Flow rate between rywell and region Flow rate from rayed region unsprayed region Flow rate from nsprayed region to sprayed region Containment leak rate to environment from sprayed 0-49 _ . Value 00 ft 0 ft 3 /min x 10 5 ft 3/min 0 ft 3/min 71,400 ft 3/min 71, 400 ft 3 /min o. 067 ft o"f (;2. ft-S/ [!QJ.q ..g,,g. percent -........... (""tr\ uncertainty* Spray fall height ....._ distribution Spray removal rate for elemental iodine (sprayed region only) Containment leak rate to unsprayed r.fi!gion G 49 SQQQRQa-11-' 4Q, lleconds -2.li .._o""r}) a t1us from sprayed region &""-..4Q 8998fl6:S 0..-4 Q, - Q:ro"m fa_@ unsprayed region 0 40 - Annulus v61umce"'-1" -

s. Flow rate from annulus to environment Annulus exhaust gas treatment system filter efficiency particulate elemental and organic iodine 15.6-59

0, "'s-ftYfll\.:'/\ 1. 96 X 10 5 ft 3 (II i'I 2000 ft 3 /min 99 percent 0 Revision 12 January, 2003 ent leak rate to environment sprayed region 40 seconds -24 hours 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months -30 days o.i&o' This revised information was marked directly into Table 15.6-12b on page 15.6-59, thereby obviating the need for this Insert page TABLE 15.6-12c LOSS-OF-COOLANT ACCIDENT PARAMETERS AND ASSUMPTIONS USED IN RADIOLOGICAL CONSEQUENCE CALCULATIONS ENGINEERED SAFETY FEATURE (ESF) LEAKAGE PATHWAY Release location Suppression pool water volume ECCS leak rate 0 -24 hours 24 -24.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months 24.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months -30 days Partition factor 15.6-60 15 gph 15 gph and 50 gpm for 30 minutes 15 gph 10 Revision 13 December, 2003 TABLE 15.6-14 CONTROL ROOM MODEL Parameter Volume 0 5 hour5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months -30 days Recirc latlon filter efficiencies pa ticulate el ental and organic iodine rate -unfiltered inleakage 0 -0.5 hour5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months 0.5 hour5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months -30 days rate -exhaust 0 -0.5 hour5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months 0.5 hour5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months -30 days b c, ""r eo.Ac._j f"' s (:) *-i. \( k 0 W"'"'$ 2 \.\. -"l \\. 0 IAI'> 'H> -7 U> h.ow-s 15.6-64 Value J..;-4.4. x 1'0 s ft l ft 3/IRiR ft 3/MR 0 2. 7 x 10 4 ft 3/min 4800 ft 3/min 1375 ft 3/min Revision 12 January, 2003

2. Offsite Doses Exclusion area (863 Meters) Low population (4,002 Meters) TABLE 15.6-15 LOSS-OF-COOLANT ACCIDENT (DESIGN BASIS RADIOLOGICAL EFFECTS Dose (Expressed as TEDE, Rem) Licensing Basis Limit (TEDE, Rem) 25 zone 25 Control Room Doses '{@ (0-30 days) .4..:.3...

15.6-65 5 Revision.12 January, 2003 MSl.2tD ----MSl.3 MSL1 '------DrywdtDMSL.1 Fia1an Pnlduct Transpolt llodal (llSIV and Contalnmant Laabge Pathways) AEGTS MSL3 CRExhaust

v t e Letter Sequence Response to RAI
TAC:MF3197, (Approved, Closed)
Initiation Request Acceptance, Acceptance Supplement Administration Questions 24 June 2014 15 January 2015 Answers 22 July 2014 4 February 2015 Results Approval Other:
MONTHYEAR2014-06-24 [Table View]

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