L-14-245, 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 FirstEnergy Nuclear Operating Co |
To: | Document Control Desk, Office of Nuclear Reactor Regulation |
References | |
L-14-245, TAC MF3197 | |
Download: ML14203A625 (93) | |
Text
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. ML13343A013).
The NRC staff requested additional information in a letter dated June 24, 2014 (Accession No. ML14162A409).
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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 0-2 hours (hrs.) 3000 cubic 0 -0.5 hour5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> (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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, leakage is reduced to 69 percent of the value used during the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> [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 <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> 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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> [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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, 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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Credit for Main Steam line leakage reduction associated with decreased pressure after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. 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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, 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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> . [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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />). [USAR pg. 15.4-29 (24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) [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 <br />0.271 hours <br />0.00161 weeks <br />3.709875e-4 months <br /> 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 <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> 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