Text

License Amendment Request 244, Radiological Accident Analysis and Discussion of Associated Technical Specification Changes, Attachment 4
	ML11252A656
Person / Time
Site:	Kewaunee
Issue date:	08/30/2011
From:	; Dominion, Dominion Energy Kewaunee
To:	; Office of Nuclear Reactor Regulation
References
	11-025A, TAC ME7110
	Download: ML11252A656 (192)
	v • d • e

Serial No. 11-025A ATTACHMENT 4 LICENSE AMENDMENT REQUEST 244:

PROPOSED REVISION TO RADIOLOGICAL ACCIDENT ANALYSIS AND CONTROL ROOM ENVELOPE HABITABILITY TECHNICAL SPECIFICATIONS RADIOLOGICAL ACCIDENT ANALYSIS AND DISCUSSION OF ASSOCIATED TECHNICAL SPECIFICATION CHANGES KEWAUNEE POWER STATION DOMINION ENERGY KEWAUNEE, INC.

Serial Number 11-025A Page 1 of 191 Radiological Accident Analyses and Discussion of Associated Technical Specification Changes Index of Contents 1.0

INTRODUCTION & BACKGROUND.............................................................................................. 5

1.1

INTRODUCTION.................................................................................................................................. 5

1.2

CURRENT LICENSING BASIS

SUMMARY

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

1.3

ANALYSIS ASSUMPTIONS & KEY PARAMETER VALUES......................................................................... 7

1.3.1

Selection of Events Requiring Reanalysis.............................................................................. 7

1.3.2

Analysis Assumptions & Key Parameter Values..................................................................... 9

2.0

PROPOSED LICENSING BASIS CHANGES.............................................................................. 15

2.1

REVISED METEOROLOGICAL X/Q VALUES FOR OFF-SITE AND CONTROL ROOM RECEPTORS................... 15

2.2

METHODOLOGY USED TO ANALYZE DOSE CONSEQUENCES USING THE RADTRAD-NAI CODE........... 15

2.3

MAXIMUM COOLANT ACTIVITY LIMITS IN TS 3.4.16............................................................................ 16

2.4

STEAM GENERATOR SECONDARY SIDE ACTIVITY LIMIT IN TS 3.7.16.................................................. 16

2.5

REQUIRE CONTROL ROOM ISOLATION PRIOR TO MOVEMENT OF RECENTLY IRRADIATED FUEL IN TS 3.7.10............................................................................................................................................. 16

2.6

ALLOW ANY CONTAINMENT PENETRATIONS TO BE OPEN UNDER ADMINISTRATIVE CONTROL (INCLUDING THE EQUIPMENT HATCH) DURING REFUELING OPERATIONS IN TS 3.9.6............................. 17

2.7

REMOVAL OF R-23 CREDIT FOR CONTROL ROOM ISOLATION............................................................. 17

2.8

DEFINITION OF DOSE EQUIVALENT I-131........................................................................................... 18

2.9

SUMMARY

OF DESIGN AND LICENSING BASIS CHANGES..................................................................... 18

3.0

RADIOLOGICAL EVENT RE-ANALYSES & EVALUATION....................................................... 25

3.1

DETERMINATION OF ATMOSPHERIC DISPERSION FACTORS (X/Q).......................................................... 27

3.1.1

Control Room X/Q................................................................................................................. 28

3.1.2

Offsite (EAB and LPZ) X/Q.................................................................................................... 34

3.2

DESIGN BASIS LOSS OF COOLANT ACCIDENT (LOCA) REANALYSIS.................................................... 35

3.2.1

LOCA Scenario Description.................................................................................................. 35

3.2.2

LOCA Source Term Definition............................................................................................... 36

3.2.3

LOCA Atmospheric Dispersion Factors................................................................................ 41

3.2.4

LOCA Containment Airborne Activity.................................................................................... 42

3.2.5

LOCA Analysis Assumptions & Key Parameter Values........................................................ 46

3.2.6

LOCA Results........................................................................................................................ 71

3.3

FUEL HANDLING ACCIDENT (FHA).................................................................................................... 72

3.3.1

FHA Scenario Description..................................................................................................... 72

3.3.2

FHA Source Term Definition................................................................................................. 73

3.3.3

FHA Release Transport......................................................................................................... 74

3.3.4

FHA Atmospheric Dispersion Factors................................................................................... 75

3.3.5

FHA Analysis Assumptions & Key Parameter Values........................................................... 75

3.3.6

FHA Analysis Results............................................................................................................ 85

3.4

STEAM GENERATOR TUBE RUPTURE ACCIDENT................................................................................ 86

3.4.1

SGTR Scenario Description.................................................................................................. 86

3.4.2

SGTR Source Term Definition............................................................................................... 87

3.4.3

SGTR Release Transport...................................................................................................... 92

3.4.4

SGTR Atmospheric Dispersion Factors................................................................................ 93

3.4.5

SGTR Key Analysis Assumptions and Inputs....................................................................... 93

3.4.6

SGTR Analysis Results....................................................................................................... 105

3.5

MAIN STEAM LINE BREAK ANALYSIS................................................................................................ 106

3.5.1

MSLB Scenario Description................................................................................................ 106

Serial Number 11-025A Page 2 of 191 3.5.2

MSLB Source Term Definition............................................................................................. 107

3.5.3

MSLB Release Transport.................................................................................................... 108

3.5.4

MSLB Atmospheric Dispersion Factors.............................................................................. 110

3.5.5

MSLB Key Analysis Assumptions and Inputs..................................................................... 112

3.5.6

MSLB Analysis Results....................................................................................................... 126

3.6

LOCKED ROTOR ACCIDENT (LRA) ANALYSIS................................................................................... 127

3.6.1

LRA Scenario Description................................................................................................... 127

3.6.2

LRA Source Term Definition................................................................................................ 128

3.6.3

LRA Release Transport....................................................................................................... 128

3.6.4

LRA Atmospheric Dispersion Factors................................................................................. 129

3.6.5

LRA Analysis Assumptions and Key Parameters............................................................... 129

3.6.6

LRA Results......................................................................................................................... 138

3.7

RCCA EJECTION ACCIDENT (REA) ANALYSIS................................................................................. 139

3.7.1

REA Scenario Description................................................................................................... 139

3.7.2

REA Source Term Definition............................................................................................... 139

3.7.3

REA Release Transport...................................................................................................... 142

3.7.4

REA Atmospheric Dispersion Factors................................................................................. 143

3.7.5

REA Analysis Assumptions and Key Parameters............................................................... 143

3.7.6

REA Analysis Results.......................................................................................................... 158

3.8

WASTE GAS DECAY TANK ANALYSIS............................................................................................... 159

3.8.1

WGDT Scenario Description............................................................................................... 159

3.8.2

WGDT Source Term Definition............................................................................................ 160

3.8.3

WGDT Release Transport................................................................................................... 160

3.8.4

WGDT Atmospheric Dispersion Factors............................................................................. 161

3.8.5

WGDT Analysis Assumptions and Key Parameters........................................................... 161

3.8.6

WGDT Analysis Results...................................................................................................... 168

3.9

VOLUME CONTROL TANK RUPTURE (VCT) ANALYSIS...................................................................... 169

3.9.1

VCT Scenario Description................................................................................................... 169

3.9.2

VCT Source Term Definition............................................................................................... 170

3.9.3

VCT Release Transport....................................................................................................... 171

3.9.4

VCT Atmospheric Dispersion Factors................................................................................. 174

3.9.5

VCT Analysis Assumptions and Key Parameters............................................................... 174

3.9.6

VCT Analysis Results.......................................................................................................... 183

4.0

ADDITIONAL DESIGN BASIS CONSIDERATIONS.................................................................. 184

4.1

RISK IMPACT OF PROPOSED CHANGES........................................................................................... 184

4.2

IMPACT UPON THE EMERGENCY PLAN............................................................................................. 187

5.0

CONCLUSIONS.......................................................................................................................... 188

6.0

REFERENCES............................................................................................................................ 189

Serial Number 11-025A Page 3 of 191 Index of Tables Table 1.3-1 Control Room Common Assumptions & Key Parameters............................................... 10

Table 1.3-2 NSS Common Assumptions & Key Parameters................................................................ 11

Table 1.3-3 Offsite Atmospheric Dispersion Factors (sec/m3)............................................................ 11

Table 1.3-4 Control Room Atmospheric Dispersion Factors............................................................... 12

Table 1.3-5 Breathing Rates.................................................................................................................... 14

Table 2.0-1 Comparative Summary of Design and Licensing Basis Changes to Radiological Event Analyses..................................................................................................................... 19

Table 3.0-1 Accident Dose Acceptance Criteria................................................................................... 26

Table 3.1-1 Line-of-Sight Horizontal Distance from Source to Receptor........................................... 32

Table 3.1-2 Direction from Receptor to Source.................................................................................... 33

Table 3.2-1 Regulatory Guide 1.183 Source Terms.............................................................................. 38

Table 3.2-2 RG 1.183 Release Phases.................................................................................................... 38

Table 3.2-3 Core Inventory and Dose Conversion Factors by Isotope............................................... 39

Table 3.2-4 Spray Removal Calculation Parameters............................................................................ 48

Table 3.2-5 Basic Data and Assumptions for LOCA............................................................................. 49

Table 3.2-6 RWST Time Dependent DF Values..................................................................................... 67

Table 3.2-7 Dose summary for a Kewaunee LOCA............................................................................... 71

Table 3.3-1 Basic Data and Assumptions for FHA............................................................................... 79

Table 3.3-2 Dose Summary for the Fuel Handling Accident Analysis................................................ 85

Table 3.4-1 Primary Coolant and Secondary Side................................................................................ 88

Table 3.4-2 Pre-accident Iodine Spike RCS Concentration................................................................. 91

Table 3.4-3 Concurrent Iodine Spike SGTR RCS Concentration......................................................... 91

Table 3.4-4 Basic Data and Assumptions for SGTR............................................................................. 96

Table 3.4-5 Time Line of Events........................................................................................................... 104

Table 3.4-6 RCS Break Flow to Affected Steam Generator................................................................ 104

Table 3.4-7 Affected Steam Generator Steam Release to Environment........................................... 104

Table 3.4-8 Intact Steam Generator Steam Release to the Environment......................................... 105

Table 3.4-9 Dose Summary for the SGTR Accident............................................................................ 105

Table 3.5-1 Concurrent Iodine Spike MSLB RCS Concentration...................................................... 108

Table 3.5-2 Basic Data and Assumptions for MSLB........................................................................... 117

Table 3.5-3 Dose Summary for the MSLB Accident........................................................................... 126

Table 3.6-1 Basic Data and Assumptions for LRA............................................................................. 132

Table 3.6-2 TEDE Results for the Locked Rotor Accident................................................................. 138

Table 3.7-1 REA Event Timing.............................................................................................................. 144

Table 3.7-2 Basic Data and Assumptions for REA............................................................................. 148

Table 3.7-3 TEDE Results for the RCCA Ejection Accident............................................................... 158

Table 3.8-1 Waste Gas Decay Tank Activity (Ci)................................................................................. 160

Table 3.8-2 Basic Data and Assumptions for WGDT.......................................................................... 164

Table 3.8-3 Dose Results for the WGDT Accident.............................................................................. 168

Table 3.9-1 Volume Control Tank Activity (Ci).................................................................................... 172

Table 3.9-2 Letdown Flow Noble Gas Concentration (Ci/gm)......................................................... 173

Table 3.9-3 Pre-Accident Iodine Spike Concentration based on 60 Ci/gm DEI............................. 173

Table 3.9-4 Basic Data and Assumptions for VCT.............................................................................. 178

Table 3.9-5 Dose Results for the VCT Accident.................................................................................. 183

Serial Number 11-025A Page 4 of 191 Index of Figures Figure 3.1-1 Kewaunee Source and Receptor Points........................................................................... 31

Figure 3.2-1 RADTRAD Model for Containment Airborne Releases................................................... 61

Figure 3.2-2 RADTRAD Model for ECCS Leakage into the Auxiliary Building.................................. 63

Figure 3.2-3 RADTRAD Model for Iodine Back-Leakage into the RWST............................................ 68

Figure 3.2-4 RADTRAD Model for Noble Gas Leakage from RWST.................................................... 69

Figure 3.3-1 RADTRAD Model for FHA.................................................................................................. 78

Figure 3.4-1 SGTR Radioactive Release Schematic............................................................................. 95

Figure 3.5-1 MSLB Radioactive Release into the Turbine Building Schematic (Design Model).... 115

Figure 3.5-2 MSLB Radioactive Release into the Auxiliary Building Schematic............................. 116

Figure 3.6-1 LRA Radioactive Release Schematic............................................................................. 131

Figure 3.7-1 RADTRAD Model for Containment Airborne Releases................................................. 146

Figure 3.7-2 RADTRAD Model for Secondary System Releases....................................................... 147

Figure 3.8-1 WGDT Radioactive Release Schematic.......................................................................... 163

Figure 3.9-1 VCT Radioactive Release Schematic.............................................................................. 175

Serial Number 11-025A Page 5 of 191 1.0 Introduction & Background 1.1 Introduction This report describes the evaluations conducted to assess off-site doses and control room habitability at Kewaunee Power Station (KPS) following postulated design basis accidents per Regulatory Guide 1.183 (Reference 1). The accident source term discussed in Reference 1 is herein referred to as the Alternative Source Term (AST).

The evaluations documented herein have employed the detailed methodology contained in RG 1.183 for use in design basis accident analyses for the AST. The results have been compared with the acceptance criteria contained either in 10 CFR 50.67 (Reference 2) or the supplemental guidance in RG 1.183.

This application, if granted, would:

Implement revised meteorological X/Q estimates (atmospheric dispersion factors) for both off-site and control room receptors from postulated accident release points

Revise the methodology used to analyze design basis dose consequences to include the RADTRAD-NAI code

Decrease Reactor Coolant Specific Activity Limit and Iodine Spike in TS 3.4.16

Decrease SG Secondary Side Activity Limit in TS 3.7.16

Require control room isolation prior to movement of recently irradiated fuel

Allow the containment penetrations to be open (including the equipment hatch) while moving recently irradiated fuel during refueling outages

Revise the TS 1.1 definition of Dose Equivalent Iodine I-131 to reference Federal Guidance Report No. 11 (FGR 11)

Require Operator action to isolate the control room within 1 hr following a Locked Rotor accident

Require Operator action to place the control room in filtered recirculation mode within 20 minutes following a Fuel Handling Accident while moving recently irradiated fuel

Serial Number 11-025A Page 6 of 191 The revised radiological dose analyses were performed with a controlled version of the computer code RADTRAD-NAI 1.1a (QA) (Reference 3). The RADTRAD computer code calculates the control room and offsite doses resulting from releases of radioactive isotopes based on user supplied atmospheric dispersion factors, breathing rates, occupancy factors and dose conversion factors. Innovative Technology Solutions of Albuquerque, New Mexico developed the RADTRAD code for the NRC. The original version of the NRC RADTRAD code was documented in NUREG/CR-6604 [Reference 4]. The Numerical Applications, Inc. (NAI) version of RADTRAD was originally derived from NRC/ITS RADTRAD, version 3.01. Subsequently, RADTRAD-NAI was changed to conform to NRC/ITS RADTRAD, Version 3.02 with additional modifications to improve usability. The RADTRAD-NAI code is maintained under NAIs QA program, which conforms to the requirements of 10 CFR 50, Appendix B.

Control Room Atmospheric Dispersion Factors were evaluated using the ARCON96 computer code (Reference 5), following the guidance of Regulatory Guide 1.194 (Reference 6). Evaluation of off-site Exclusion Area Boundary (EAB) and Low Population Zone (LPZ) Atmospheric Dispersion Factors was performed with a controlled version of the Dominion computer code PAVAND (Reference 7) which is a Dominion variant of the NRC PAVAN code. The EAB and LPZ X/Q values were developed following the guidance of Regulatory Guide 1.145 (Reference 8).

1.2 Current Licensing Basis Summary The current design basis radiological analyses that appear in the KPS Updated Safety Analysis Report (USAR) consist of assessments of the following events:

1. Loss of Coolant Accident

2. Fuel Handling Accident

3. Steam Generator Tube Rupture

4. Main Steam Line Break

5. Locked Rotor Accident

Serial Number 11-025A Page 7 of 191

6. Rod Control Cluster Assembly (RCCA) Ejection Accident

7. Waste Gas Decay Tank Failure

8. Volume Control Tank Rupture (Atmospheric Release)

The analyses of record for the above events were previously docketed in Kewaunee Power Station (KPS) Amendment No. 166, issued March 17, 2003 (Reference 10),

which implemented the AST; and Amendment No. 172, issued February 27, 2004 (Reference 11), which implemented a stretch power uprate to 1772 megawatt thermal (MWt). These approved radiological accident analyses used the analytical methods and assumptions outlined in RG 1.183. By letter dated January 30, 2006 (Reference 12), as supplemented by letter dated January 23, 2007 (Reference 13), DEK requested an amendment to modify the radiological accident analyses and associated TS. This amendment incorporated TS changes to compensate for the higher control room emergency zone (CREZ) unfiltered in-leakage measured during the American Society for Testing and Materials (ASTM) E741 (tracer gas) leakage test conducted in December 2004. The NRC approved this proposed amendment as KPS License Amendment 190 on March 8, 2007 (Reference 14).

1.3 Analysis Assumptions & Key Parameter Values 1.3.1 Selection of Events Requiring Reanalysis Kewaunee Power Station has received approval for full implementation of the AST (as defined in Section 1.2.1 of Reference 1).

To support the licensing basis and plant operation changes discussed in Section 2.0 of this application, the following accidents were reanalyzed employing the guidance of RG 1.183:

Loss of Coolant Accident (LOCA),

Fuel Handling Accident (FHA),

Steam Generator Tube Rupture (SGTR) Accident,

Serial Number 11-025A Page 8 of 191

Main Steam Line Break (MSLB) Accident,

Locked Rotor Accident (LRA) and

Rod Control Cluster Assembly (RCCA) Ejection Accident (REA).

The Waste Gas Decay Tank (WGDT) failure and Volume Control Tank (VCT) rupture (Atmospheric Release) radiological analyses are also being updated to reflect revised X/Q values determined in Section 3.1 of this application. Both analyses demonstrate acceptable dose to control room operators without credit of control room emergency ventilation or isolation as well as acceptable results to the EAB under Branch Technical Position (BTP) ETSB 11-5, Rev 0 (Reference 19).

The proposed licensing basis and plant operational changes are discussed in Section 2.0. These changes require appropriate changes to the KPS Technical Specifications, which are also described in Section 2.0 of this report. The key changes considered are listed below:

a. Revise definition of Dose Equivalent I-131 in Section 1.1 of the Technical Specifications to reference Federal Guidance Report No. 11 (Reference 15) as the source of thyroid committed dose equivalent (CDE) dose conversion factors.

b. Revise Technical Specification 3.4.16, to decrease the RCS activity limits to 0.1 Ci/gm DE I-131 and 16.4 Ci/gm DE Xe-133.

c. Revise Technical Specification 3.4.16, to decrease the pre-existing iodine spike limit from 20 Ci/gm DE I-131 to 10 Ci/gm DE I-131.

d. Revise Technical Specification 3.7.16, to decrease the SG bulk liquid concentration limit from 0.1 Ci/gm to 0.05 Ci/gm DE I-131.

e. Revise Technical Specification 3.7.10, to require isolation of the control room prior to movement of recently irradiated fuel.

f. Revise Technical Specification 3.9.6, to allow ANY containment penetrations to be open under Administrative Control (including the equipment hatch) during Refueling Operations.

Serial Number 11-025A Page 9 of 191

g. Revise 3.3.7 to remove Actions and Surveillance Requirements associated with R23 instrumentation.

h. Revise the appropriate TS Bases Sections to reflect the above listed changes in accordance with the KPS Bases Control Program as described in Section 5.5.12 of the Technical Specifications.

It can be concluded from the discussion above that implementing the revised X/Q values, in conjunction with the proposed plant operational changes, will require reanalysis of the LOCA, FHA, SGTR, MSLB, LRA, REA, WGDT and VCT. Sections 3.2 through 3.9, respectively, provide detailed descriptions of the re-analyses for these events.

1.3.2 Analysis Assumptions & Key Parameter Values This section describes the general analysis approach and presents analysis assumptions and key parameter values that are common to all the accident analyses.

The dose analyses documented in this application employ the Total Effective Dose Equivalent (TEDE) calculation method, as specified in RG-1.183 for AST applications.

The Total Effective Dose Equivalent (TEDE) is determined at the Exclusion Area Boundary (EAB) for the worst 2-hour interval. TEDE for individuals at the Low Population Zone (LPZ) and for the KPS Control Room personnel are calculated for the assumed 30-day duration of the event.

The TEDE concept is defined to be the Deep Dose Equivalent, DDE, (from external exposure) plus the Committed Effective Dose Equivalent, CEDE, (from internal exposure). In this manner, TEDE assesses the impact of all relevant nuclides upon all body organs, in contrast with the previous single, critical organ (thyroid) concept for assessing internal exposure. CEDE dose conversion factors were taken from Table 2.1 of Federal Guidance Report 11 (Reference 15) per Section 4.1.2 of Regulatory Guide 1.183. The DDE is nominally equivalent to the Effective Dose Equivalent (EDE) from external exposure if the whole body is irradiated uniformly. Since this is a reasonable assumption for submergence exposure situations, EDE is used in lieu of DDE in

Serial Number 11-025A Page 10 of 191 determining the contribution of external dose to the TEDE. EDE dose conversion factors were taken from Table III.1 of Federal Guidance Report 12 (Reference 16) per Section 4.1.4 of Regulatory Guide 1.183.

There are a number of analysis assumptions and plant features that are used in the analysis of all of the events. These assumptions and features are presented in Tables 1.3-1 through 1.3-5.

Table 1.3-1 Control Room Common Assumptions & Key Parameters Assumption / Parameter Value Control Room Effective Volume 127,600 ft3 Control Room Intake Flow Rate prior to Isolation 2750 cfm Unfiltered Control Room Inleakage 800 cfm Emergency Ventilation System Recirculation Flow Rate 2500 cfm + 10%

Response Time for Control Room to Isolate upon Receipt of a Safety Injection (SI) Signal 10 seconds Delay to Control Room Post Accident Recirculation Mode (CRPARS) operation following Receipt of a SI Signal

10 sec Delay to Diesel Start-up

63 sec Delay to Sequence Diesel to CRPARS

60 sec Delay to Open Recirc damper 133 seconds Control Room Filter Efficiencies Elemental:

90%

Organic:

90%

Particulate: 99%

Control Room Wall Thickness:

>1.5 feet Concrete Control Room Ceiling Thickness:

>1.5 feet Concrete Control Room Occupancy Factors 0 - 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 24 - 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months 96 - 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months 1.0 0.6 0.4

Serial Number 11-025A Page 11 of 191 Table 1.3-2 NSS Common Assumptions & Key Parameters Assumption / Parameter Value Internal Reactor Containment Vessel Free Volume 1.32E6 ft3 Shield Building Free Volume 3.74E5 ft3 Shield Building Wall Thickness:

2.5 ft Concrete Shield Building Dome Thickness:

2.0 ft Concrete Internal Containment Inner Radius:

52.5 ft Shield Building Inner Radius 57.5 ft Table 1.3-3 Offsite Atmospheric Dispersion Factors (sec/m3)

Location / Duration

/Q (sec/m3)

Exclusion Area Boundary (EAB=1200 m radius)

All Release Points 0 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 1.76E-04 Low Population Zone (LPZ=3 mile*)

All Release Points 0 - 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 8 - 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 1 - 4 days 4 - 30 days 3.36E-05 2.37E-05 1.12E-05 3.94E-06

Conservatively calculated at 2 miles

Serial Number 11-025A Page 12 of 191 Table 1.3-4 Control Room Atmospheric Dispersion Factors Source / Accident / Duration Control Room Intake /Q (sec/m3)

Isolated CR Worst

In-leakage X/Q (sec/m3)

Reactor Building Stack Exhaust (LOCA,REA&FHA) 0 - 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 2 - 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 8 - 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 24 - 96 hour0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months 96 - 720 hour0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months 4.88E-03 3.51E-03 1.37E-03 1.12E-03 9.41E-04 3.97E-03 2.95E-03 1.11E-03 8.89E-04 7.87E-04 Containment / Shield Building (LOCA&REA) 0 - 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 2 - 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 8 - 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 24 - 96 hour0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months 96 - 720 hour0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months 1.84E-03 1.23E-03 5.03E-04 4.22E-04 3.50E-04 1.74E-03 1.16E-03 4.70E-04 4.02E-04 3.28E-04 Auxiliary Building Stack Exhaust (LOCA,REA,FHA,MSLB,WGDT&VCT) 0 - 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 2 - 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 8 - 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 24 - 96 hour0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months 96 - 720 hour0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months 3.67E-03 2.83E-03 1.11E-03 7.34E-04 5.64E-04 2.90E-03 2.26E-03 8.79E-04 5.80E-04 4.47E-04 Containment Equipment Hatch (FHA) 0 - 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 2 - 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 8 - 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 24 - 96 hour0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months 96 - 720 hour0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months 3.41E-03 2.88E-03 1.22E-03 9.71E-04 7.66E-04 4.58E-03 3.88E-03 1.64E-03 1.32E-03 1.07E-03 Fuel Area Roll-up Door (FHA) 0 - 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 2 - 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 8 - 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 24 - 96 hour0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months 96 - 720 hour0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months 1.44E-03 1.26E-03 5.27E-04 4.23E-04 3.56E-04 1.53E-03 1.35E-03 5.61E-04 4.51E-04 3.83E-04

Serial Number 11-025A Page 13 of 191 Table 1.3-4 Control Room Atmospheric Dispersion Factors Source / Accident / Duration Control Room Intake /Q (sec/m3)

Isolated CR Worst

In-leakage X/Q (sec/m3)

A Steam Generator PORV (MSLB,SGTR,LRA&REA) 0 - 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 2 - 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 8 - 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 24 - 96 hour0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months 96 - 720 hour0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months 2.24E-03 1.90E-03 7.69E-04 6.37E-04 5.19E-04 2.46E-03 2.13E-03 8.60E-04 6.96E-04 5.81E-04 A Steam Generator Safeties NotUsed

BoundedbyASG

PORV NotUsed

BoundedbyASG

PORV A Steam Generator Dumps NotUsed

BoundedbyASG

PORV NotUsed

BoundedbyASG

PORV B Steam Generator PORV (MSLB,SGTR,LRA&REA)

0 - 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 2 - 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 8 - 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 24 - 96 hour0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months 96 - 720 hour0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months 3.96E-02*

3.20E-02*

1.21E-02 1.01E-02 8.58E-03 2.92E-02*

2.34E-02*

8.67E-03 6.97E-03 6.41E-03 B Steam Generator Safeties NotUsed

BoundedbyBSG

PORV NotUsed

BoundedbyBSG

PORV B Steam Generator Dumps NotUsed

BoundedbyBSG

PORV NotUsed

BoundedbyBSG

PORV

The most significant pathway of inleakage to the Control Room is through doorway penetrations in communication with the Turbine Building. The worst in-leakage X/Q is the highest X/Q from the following possible intake points to the Turbine Building: TB Fan Room West Louvers, TB Fan Room East Louvers, and TB Roll-up Door.

The value displayed can be and was divided by 5 for use in the SGTR and/or LRA dose analyses.

This reduction by a factor of 5 was permitted due to the steam exhaust vertical velocity exceeding the 95th percentile wind speed at the release elevations for the SGTR and LRA. Division by 5 is only applicable for the 0-2 hour interval for the SGTR and the 0-2 hour and 2-8 hour intervals for the LRA and SGTR. Justification for this reduction by a factor of 5 is given in Section 3.4.5.3 (SGTR) and Section 3.6.5.3 (LRA) and the results are shown in Tables 3.4-4 (SGTR) and 3.6-1 (LRA).

Serial Number 11-025A Page 14 of 191 Table 1.3-5 Breathing Rates Location / Duration (m3/sec)

Offsite (EAB & LPZ) 0 - 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 8 - 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 24 - 720 hour0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months 3.5E-04 1.8E-04 2.3E-04 Control Room 0 - 720 hour0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months 3.5E-04

Serial Number 11-025A Page 15 of 191 2.0 Proposed Licensing Basis Changes This section provides a summary description of the key proposed licensing basis changes that are justified with the revised KPS AST analyses contained within this attachment.

2.1 Revised Meteorological X/Q Values for Off-site and Control Room Receptors This analysis supports a request to revise the design basis accident atmospheric dispersion factor (X/Q) values for KPS. Atmospheric dispersion factors are significant inputs in assessments performed to demonstrate compliance with 10 CFR Part 50. The determinations of off-site and control room X/Q values were made pursuant to the guidance of Regulatory Guides 1.145 and 1.194, respectively. After approval of this licensing basis change, the X/Q used in evaluating the consequences of design basis accidents will become the official and documented values in the USAR.

2.2 Methodology Used to Analyze Dose Consequences Using the RADTRAD-NAI Code This analysis supports a request to revise the methodology used to evaluate design basis accident dose consequences to include using the RADTRAD-NAI code. Currently approved analyses-of-record were developed by Westinghouse using proprietary codes and methods. Dominion re-analyses using RADTRAD-NAI will replace the existing Westinghouse methodology used in evaluating the dose consequences of design basis accidents and continue to follow the guidance of RG 1.183. This license amendment application is made pursuant to the requirements of 10 CFR 50 which specifies that a revision to the methodology described in the Updated Safety Analysis Report, such as the design basis radiological consequence analyses, shall be submitted for approval.

The proposed changes for radiological events have been analyzed and result in acceptable consequences, meeting the criteria as specified in 10 CFR 50.67 and RG 1.183.

Serial Number 11-025A Page 16 of 191 2.3 Maximum Coolant Activity Limits in TS 3.4.16 The limits on maximum primary coolant activity ensure that the analyzed post-accident dose consequences of design basis accidents meet the limits specified in GDC 19 and 10 CFR 50.67. The proposed change involves decreasing the reactor coolant specific activity limits to < 0.1 Ci/gram DE I-131 and <16.4 Ci/gram DE Xe-133. The DE Xe-133 limit is set to be consistent with the level of fuel damage equivalent to 0.1 Ci/gram DE I-131 (i.e., ~0.03% failed fuel). The pre-existing iodine spike threshold is also being reduced to <10 Ci/gram DE I-131, commensurate with the limit reduction in reactor coolant specific activity. The applicable accidents analyzed for this spike ensure control room and off-site post-accident doses are within the acceptance criteria of GDC-19 and a fraction of 10 CFR 50.67 limits.

2.4 Steam Generator Secondary Side Activity Limit in TS 3.7.16 In conjunction with the proposed decrease in primary coolant activity, a lower secondary side activity limit of < 0.05 Ci/gram DE I-131 is also proposed. The decreases in Technical Specification activity limits were necessary to result in acceptable dose consequences following a radiological event. The proposed changes in the primary coolant activity (Section 2.0.C) and the secondary side activity, coupled with the methods and assumptions specified by RG 1.183, result in estimated accident dose consequences meeting the acceptance criteria of 10 CFR 50.67 and RG 1.183.

2.5 Require control room isolation prior to movement of recently irradiated fuel in TS 3.7.10 The Technical Specification Refueling Operations Requirements define criteria necessary to result in acceptable dose consequences following a fuel handling accident.

The proposed change to require control room isolation during movement of recently irradiated fuel is necessary to achieve acceptable control room occupant doses.

Serial Number 11-025A Page 17 of 191 2.6 Allow ANY containment penetrations to be open under Administrative Control (including the equipment hatch) during Refueling Operations in TS 3.9.6 The analysis of the consequences from a Fuel Handling Accident (FHA) in either the containment or fuel storage pool, use the accident criteria specified by RG 1.183 and assume open penetrations in the containment and/or the fuel storage pool area. The analysis is modeled for the worst case release scenario (e.g., highest control room X/Q and bounding off-site X/Qs with a complete release of fuel bundle radioactivity over a 2-hour duration directly to the atmosphere). The resulting dose consequences continue to meet the acceptance criteria of 10 CFR 50.67 and RG 1.183.

Any open penetrations to the containment or fuel storage pool during movement of recently irradiated fuel will be identified and administratively controlled to ensure personnel and equipment are designated to promptly close the penetration(s).

Administrative controls include:

Appropriate personnel are aware that penetrations are open, A specified individual(s) is designated and available to close each penetration following a fuel handling event, and Any obstruction(s) (e.g., cables and hoses) that could prevent closure of any penetration can be quickly removed.

2.7 Removal of R-23 Credit for Control Room Isolation Credit for the Control Room Ventilation Intake radiation monitor R-23, which provides control room isolation, is being removed. The R-23 system is not safety grade and consists of a single radiation monitor. In addition, the isolation signal generated by R-23 is only a partial signal that will not assure closure of all control room inlet and outlet ventilation dampers to provide complete control room isolation. Full control room isolation requires actions by the operator to close minor dampers that are not included in the isolation logic. Only the current Fuel Handling Accident (FHA) and Locked Rotor Accident (LRA) events use and credit the R-23 system for control room isolation. These

Serial Number 11-025A Page 18 of 191 analyses currently rely on the assumptions that Operations will take appropriate actions to isolate the control room if R-23 fails to perform its isolation function.

Removing credit for R-23 requires an alternative means to ensure control room isolation in the event of a FHA or LRA. The proposed new FHA requires that the control room be isolated prior to moving recently irradiated fuel, so therefore, R-23 is no longer required for that accident. As discussed in Section 3.6, a proposed operator action will be required within one hour following a LRA to isolate the control room. One hour is sufficient time for the operator to identify the accident, take necessary emergency steps in response to the accident, and direct action to isolate the control room and start the control room post accident recirculation system (CRPARS).

Since R-23 is not longer credited to perform any safety function, TS 3.3.7 will be modified to remove all TS Actions and Surveillance Requirements associated with R-23 instrumentation.

2.8 Definition of Dose Equivalent I-131 A change to the Technical Specification Definition of Dose Equivalent Iodine I-131 is proposed to reference Federal Guidance Report No. 11 (FGR 11), "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion," 1989, as the source of thyroid CDE dose conversion factors (Reference 15).

2.9 Summary of Design and Licensing Basis Changes This Section provides a comparative summary of the current design and licensing basis and the proposed changes. The summary is listed in Table 2.0-1. A detailed discussion of the changes, including the reasons for the changes, can be found in Section 3.

Serial Number 11-025A Page 19 of 191 The existing analyses for the radiological events, as listed in Section 1.2, were performed at various times using different codes and/or hand calculations. The common element for these events is the use of a single bounding control room X/Q and a single set of off-site dispersion factors (X/Q) to assess resulting radiological consequences. The proposed amendment utilizes new estimates of control room and off-site (EAB and LPZ) dispersion factors using the guidance provided in Regulatory Guides 1.145 and 1.194, and supporting documents. Additionally, a comprehensive design basis validation for all inputs and assumptions used in each radiological analysis was performed. This accounts for differences in some of the parameters listed in Table 2.0-1.

Table 2.0-1 Comparative Summary of Design and Licensing Basis Changes to Radiological Event Analyses Parameter Current Basis Proposed Basis Alternate Source Term RCS Technical Specification Limits 1.0 Ci/gm DE I-131 AND 595 Ci/gm DE Xe-133 0.1 Ci/gm DE I-131 AND 16.4 Ci/gm DE Xe-133 RCS Technical Specification Gross Gamma Concentrations USAR Table D.4-1 (based on 1% fuel defects)

Table 3.4-1 (based on fuel defects equivalent to 0.1 Ci/gm DE I-131)

RCS Technical Specification Iodine Concentrations Isotope I-131 I-132 I-133 I-134 I-135 1.0 Ci/gm DE I-131 Conc.

(Ci/gm) 7.80E-01 7.93E-01 1.16E+00 1.61E-01 6.37E-01 0.1 Ci/gm DE I-131 Conc.

(Ci/gm) 7.82E-02 7.97E-02 1.17E-01 1.62E-02 6.40E-02 Secondary Side Technical Specification Limit

< 0.1 Ci/gm DE I-131

< 0.05 Ci/gm DE I-131 Pre-accident Iodine Spike 20 Ci/gm DE I-131 10 Ci/gm DE I-131

Serial Number 11-025A Page 20 of 191 Table 2.0-1 Comparative Summary of Design and Licensing Basis Changes to Radiological Event Analyses Parameter Current Basis Proposed Basis Iodine Appearance Rates Isotope I-131 I-132 I-133 I-134 I-135 Conc.

(Ci/min) 0.301 0.788 0.519 0.319 0.377 Conc.

(Ci/min) 0.030 0.079 0.052 0.032 0.038 Dose Conversion Factors ICRP30 FGR 11 and 12 Offsite Dose Historical X/Qs (unknown basis)

Revised X/Qs (based on RG 1.145 and PAVAND)

Offsite Breathing Rates 0 - 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 8 - 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 24 - 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months 3.47E-04 1.75E-04 2.32E-04 3.5E-04 1.8E-04 2.3E-04 Control Room Unfiltered Inleakage (cfm) 800 (based on SI signal)

OR 1500 (based on R-23 signal) 800 credit for R-23 has been removed X/Qs Murphy & Campe ARCON96 (listed in Table 1.3-4)

Breathing Rate 3.47E-04 3.5E-04 Loss-of Coolant Accident (Section 3.2)

Iodine Chemical Form in the Sump (%)

100% Elemental 97% Elemental 3% Organic Containment Sump Volume (gal) 315,000 311,000 Containment Spray Duration (hr) 0.917 0.91

Serial Number 11-025A Page 21 of 191 Table 2.0-1 Comparative Summary of Design and Licensing Basis Changes to Radiological Event Analyses Parameter Current Basis Proposed Basis Containment spray Removal Coefficient (hr-1)

Elemental Particulate 20 4.5 15 2.8 Natural deposition (hr-1) 0.1 Powers Model set at the 10th percentile ECCS Iodine Airborne Evolution (%)

0-3 hour

>3 hour 10 1

10 10 RWST Iodine Airborne Evolution 1%

DF=100 RWST Backleakage modeling Section 3.2.5.5 Fuel Handling Accident (Section 3.3)

Unfiltered Inleakage after control room isolation (cfm) 1500 800 Credited operator action None Control room is isolated prior to movement of recently irradiated fuel.

Operator action to place control room in filtered recirculation mode within 20 minutes of FHA.

Control Room Configuration while Moving Recently Irradiated Fuel Normal Isolated prior to movement

Serial Number 11-025A Page 22 of 191 Table 2.0-1 Comparative Summary of Design and Licensing Basis Changes to Radiological Event Analyses Parameter Current Basis Proposed Basis Steam Generator Tube Rupture Accident (Section 3.4)

Release termination of Primary to Secondary Leakage for Intact Steam Generators (hours) 24 29 Duration of break flow and discharge from Affected Steam Generator (min) 30 55 Operator Action to close Affected SG PORV (min) 30 55 Total Break Flow (lbm) 0-30 min:

154,900 0-55 min:

282,100 Condenser as a release pathway Credited Not Credited Iodine Spike 500 335 Iodine Spike duration 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 8 hours Main Steam Line Break Accident (Section 3.5)

Safety Injection Signal (sec) 0

<3 Action to Align RHR (hr) 24 29 Release to Environment (hr)

Unaffected SG Affected SG 0 - 24 0 - 29

Serial Number 11-025A Page 23 of 191 Table 2.0-1 Comparative Summary of Design and Licensing Basis Changes to Radiological Event Analyses Parameter Current Basis Proposed Basis Pre-accident spike Concurrent spike 72 72 69.2 8

Operator Action - close Affected SG MSIV (hr)

NA 8

Release of Initial Mass in Faulted Generator (min) 2 10 Accident-Initiated (Concurrent) Spike Duration (hr) 4 8

Duration of Primary to Secondary Leakage for Affected Steam Generator 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 69.2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Duration of Primary to Secondary Leakage for Intact Steam Generators 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 29 hours Locked Rotor Accident (Section 3.6)

Failed Fuel Following the Accident (%)

50 25 Steam Generator Liquid Mass (lbm/SG) 0 - 30 minutes 87,000 84,000 Control Room Isolation (min) 10.67 60 RCCA Ejection Accident (Section 3.7)

Safety Injection Signal (sec) 52.5 240 Steam Generator Liquid Mass (lbm/SG) 87,000 84,000

Serial Number 11-025A Page 24 of 191 Table 2.0-1 Comparative Summary of Design and Licensing Basis Changes to Radiological Event Analyses Parameter Current Basis Proposed Basis WGDT Accident (Section 3.8)

Dose Consequence Multiplier (Method to adjust cycle activity to account for changes in operating conditions and fuel management variations) 1.1 1.12 Release Duration (min) 5 120 Release Rate (%/day) 1.99E+05 8.289E+03 Control Room Isolation (min) 0.5 30 VCT Accident (Section 3.9)

Dose Consequence Multiplier (Method to adjust cycle activity to account for changes in operating conditions and fuel management variations) 1.1 1.12 Control Room Unfiltered Inleakage (cfm) 0 200

Serial Number 11-025A Page 25 of 191 3.0 Radiological Event Re-Analyses & Evaluation As documented in Section 1.3.1, this application involves the reanalysis of the design basis radiological analyses for the following accidents:

Loss-of-Coolant Accident (LOCA)

Fuel Handling Accident (FHA)

Steam Generator Tube Rupture (SGTR) Accident

Main Steam Line Break (MSLB) Accident

Locked Rotor Accident (LRA)

Rod Control Cluster Assembly (RCCA) Ejection Accident (REA)

Waste Gas Decay Tank (WGDT) Failure Accident

Volume Control Tank (VCT) Rupture Accident The calculated radiological consequences are compared with the limits provided in 10 CFR 50.67(b)(2), and as clarified per the additional guidance in RG-1.183 for events with a higher probability of occurrence.

New atmospheric dispersion factors (X/Qs) have been calculated. New control room X/Qs were calculated using the ARCON96 code (Reference 5) and guidance from Regulatory Guide 1.194 (Reference 6). The most limiting onsite accident X/Qs were selected from source/receptor pairs which included those potentially associated with single failure, loss of offsite power, control room pre-isolation and post-isolation intake and inleakage points. The offsite Exclusion Area Boundary (EAB) and Low Population Zone (LPZ) X/Qs were calculated using the PAVAND code (Reference 7) and guidance from Regulatory Guide 1.145 (Reference 8).

Dose calculations are performed at the EAB for the worst 2-hour period, and for the LPZ and KPS Control Room for the duration of the accident (30 days). All of the radiological dose consequence calculations were performed with the RADTRAD-NAI computer code system (Reference 3). The dose acceptance criteria that apply are provided in Table 3.0-1.

Serial Number 11-025A Page 26 of 191 Table 3.0-1 Accident Dose Acceptance Criteria Accident or Case Control Room(1)

EAB & LPZ Design Basis LOCA 5 rem TEDE 25 rem TEDE Steam Generator Tube Rupture Fuel Damage or Pre-accident Spike Coincident Iodine Spike 5 rem TEDE 5 rem TEDE 25 rem TEDE 2.5 rem TEDE(2)

Main Steam Line Break Fuel Damage or Pre-accident Spike Coincident Iodine Spike 5 rem TEDE 5 rem TEDE 25 rem TEDE 2.5 rem TEDE(2)

Locked Rotor Accident 5 rem TEDE 2.5 rem TEDE(2)

RCCA Ejection Accident 5 rem TEDE 6.3 rem TEDE(2)

Fuel Handling Accident 5 rem TEDE 6.3 rem TEDE(2)

Waste Gas Decay Tank Failure 5 rem TEDE 0.5 rem WB(3)

Volume Control Tank Rupture 5 rem TEDE 0.5 rem WB(3)

(1)

Based on 10CFR50.67 and 10 CFR 50, Appendix A, GDC 19 (2)

Reduced from 10 CFR 50.67 criteria in accordance with RG 1.183 for higher probability events.

(3)

Current licensing basis

Serial Number 11-025A Page 27 of 191 3.1 Determination of Atmospheric Dispersion Factors (X/Q)

A comprehensive evaluation of X/Q values applicable to the radiological events listed in Section 1.3.1 has been performed. Release points for each accident scenario were identified and paired with possible receptor locations to determine the most limiting X/Q values. The most limiting X/Q values were used to model the dose consequences.

Onsite source/receptor pairs were evaluated using the qualified and tested ARCON96 code (Reference 5) while the offsite source/receptor pairs to the EAB and LPZ were evaluated with a controlled version of the Dominion computer code PAVAND (Reference 7) which is a Dominion variant of the NRC PAVAN code.

of this Attachment includes a computer file on CDROM which contains the site meteorological data collected over the years 2002-2006 and used as the primary input in the calculation of the atmospheric dispersion factors. The meteorological data for KPS collected over this period were collected and processed in accordance with the standards described in RG 1.23 (Reference 18). Additionally, Enclosure 1 also includes the ARCON96 and PAVAND input files that were used in the calculation of the control room and offsite X/Q values.

The meteorological data is hourly as described in Regulatory Guide 1.23. This data has been reviewed by meteorologists for missing or anomalous observations, instrumentation problems, and trends indicative of local effects such as building wakes and excessive vegetation effects. The data meets the requirement of Regulatory Guide 1.23 for annual joint recovery rates of at least 90%.

During the review of the meteorological data, the meteorologists observed that there was a change in the distribution of the atmospheric stability classes in the data during early January of 2005. After January 2005, the occurrence of extremely and moderately unstable stability classes increased from the distribution observed from the previous three years of data. At the same time, the occurrence of slightly stable stability classes decreased. An effort was made to determine the cause of this shift in stability class distribution. During January of 2005, the Kewaunee plant process computer was

Serial Number 11-025A Page 28 of 191 replaced. The algorithm used to calculate the stability class was examined. The algorithm was found to comply with requirements and methods. The stability classes since Jan 2005 were compared to available Point Beach data and they matched well.

Point Beach is located just a few miles south of KPS. The conclusion reached was that the change in stability class distribution was tied to the replacement of the plant process computer. However, no conclusion could be reached on whether the stability class distribution, before the plant process computer change, was necessarily incorrect.

Intuitively, an increase in the percentage of highly unstable wind conditions should cause the resulting atmospheric dispersion factors to be smaller. Based on the stability class distribution, it was believed that use of only the final 2 years of data would result in smaller X/Q values. Use of only the first 3 years of data could be overly conservative.

Since the last two years of data meet quality standards and compare favorably to data recorded for the same period at Point Beach, the use of only the first 3 years of data, which contain a larger distribution of stable atmospheric conditions for unknown reasons, did not seem appropriate. Therefore, the meteorological data for all 5 years were used and are believed to be appropriate and conservative.

3.1.1 Control Room X/Q Control room X/Qs are calculated for both ventilation intake and potential inleakage receptor points to the control room and are listed in Table 1.3-4. Figure 3.1-1 provides a relative scaled drawing of the KPS building orientation and control room location showing all identified release points and receptors. The control room envelope is physically within the Auxiliary Building with ingress/egress doors into both the Auxiliary and Turbine Buildings.

DEK believes the primary source of inleakage into the control room occurs through the ingress/egress doors. This conclusion is based on the following:

a. In December of 2004, tracer gas tests were performed to measure the unfiltered in-leakage into the KPS control room. Based on observations and measurements

Serial Number 11-025A Page 29 of 191 obtained during those tests, the ingress/egress doors appeared to be the most viable source of inleakage when the control room is isolated.

b. The isolation dampers in the normal and alternate control room intakes are bubble-tight dampers. Due to the nature of their design, no inleakage is expected to occur past these dampers when closed.

c. Due to multiple areas within the Auxiliary Building being under suction by the Special Ventilation System, some directly adjacent to the control room boundary, the primary pathway and source of inleakage through the control room doors is considered to be from the turbine building.

Due to the facts above, the most viable intake to the control room when the normal control room intake is isolated is from the Turbine Building through the ingress/egress doors. Various intake points to the Turbine Building were considered as receptor locations and are shown in Figure 3.1-1. These locations are: Turbine Building Fan Room West Louver, Turbine Building Fan Room East Louver, and the Turbine Building Roll-up Door. No credit is taken for dilution within the large Turbine Building volume or additional dispersion within the Turbine Building as the contaminants travel from the intake point to the likely control room inleakage doorways. In essence, the intake into the Turbine Building is being conservatively treated very similar to a ventilation duct leading directly to the control room.

As a result of the analyses documented in this LAR, the alternate control room intake will be restricted from use. This restriction is required because of the X/Q that would result due to the close proximity of the alternate intake to various release points; one of which is < 10 m from the alternate intake. Administrative controls will be in place to assure the alternate control room intake is closed and prohibit its use during normal operation, following an accident, or while moving recently irradiated fuel.

Control room X/Q values for the source/receptor pairs address the most viable locations and limiting accident cases, including those potentially associated with single failure and loss of offsite power. The ARCON96 input source-to-receptor distances were the

Serial Number 11-025A Page 30 of 191 shortest horizontal (X-Y) distance between the release point and intake, regardless of intervening buildings (i.e., source to receptor taunt-strings or elevation differences were not considered). Table 3.1-1 and 3.1-2 provide the distances and angles for each source to receptor combination.

In accordance with the guidance of RG 1.194, the buoyant plume rise associated with energetic releases from steam relief values or atmospheric steam dumps can be credited if (1) the release is uncapped and vertical, and (2) the time-dependent vertical velocity exceeds the 95th percentile wind speed, at the release point height, by a factor of 5. Justification for crediting buoyant plume rise is given in Section 3.4.5.3 (SGTR) and Section 3.6.5.3 (LRA).

Serial Number 11-025A Page 31 of 191 Figure 3.1-1 Kewaunee Source and Receptor Points

Serial Number 11-025A Page 32 of 191 Table 3.1-1 Line-of-Sight Horizontal Distance from Source to Receptor (meters)

Elev.*

21.69 m 17.91m 17.91 m 3.51 m Control Room Intake TB Fan Room West Louver TB Fan Room East Louver TB Roll-up Door 51.35 m Rx Bldg Stack 17.05 16.93 21.67 33.61 37.13 m Shield Bldg 14.58 12.34 18.45 29.83 27.89 m Aux Bldg Stack 39.60 44.89 44.20 53.23 4.27 m Equipment Hatch 39.60 34.61 41.46 50.39 3.51 m Fuel Area Roll-up Doors1 64.55 62.98 68.83 80.44 12.60 m SG A PORV 53.35 50.87 57.14 68.33 22.35 m SG A Dump 57.93 56.53 62.27 73.95 12.60 m SG A Safeties2 53.79 51.67 57.83 69.15 23.34 m SG B PORV 12.06 12.81 16.84 28.82 25.83 m SG B Dumps3 24.81 30.56 29.16 38.00 23.34 m SG B Safeties4 13.25 13.46 17.90 29.85

Above grade (meters) 1 Fuel Area Roll-up Door #2 (south) to all receptors 2 Safety #2 to CR Intake, TB FR East Louver; Safety #1 to TB FR West Louver, TB Roll-Up Door 3 Dump #1 (South) for all receptors 4 Safety #1 to all receptors RECEPTORS RELEASE POINTS

Serial Number 11-025A Page 33 of 191 Table 3.1-2 Direction from Receptor to Source (degrees true North)

Control Room Intake TB Fan Room West Louver TB Fan Room East Louver TB Roll-up Door Rx Bldg Stack 286.4° 308.7° 293.9° 296.1° Shield Bldg 273.3° 284.4° 279.1° 284.2° Aux Bldg Stack 349.4° 354.8° 346.0° 336.7° Equipment Hatch 246.0° 252.6° 252.9° 263.0° Fuel Area Roll-up Doors1 282.0° 287.8° 284.6° 286.9° SG A PORV 273.2° 279.9° 277.0° 280.9° SG A Dump 283.1° 289.6° 285.9° 288.2° SG A Safeties2 277.4° 281.6° 280.8° 282.2° SG B PORV 289.7° 320.2° 298.4° 299.1° SG B Dumps3 356.2° 2.8° 349.9° 336.0° SG B Safeties4 286.2° 314.6° 295.3° 297.2° 1 Fuel Area Roll-up Door #2 (south) to all receptors 2 Safety #2 to CR Intake, TB FR East Louver; Safety #1 to TB FR West Louver, TB Roll-Up Door 3 Dump #1 (South) for all receptors 4 Safety #1 to all receptors RECEPTORS RELEASE POINTS

Serial Number 11-025A Page 34 of 191 3.1.2 Offsite (EAB and LPZ) X/Q The Exclusion Area Boundary (EAB) and Low Population Zone (LPZ) atmospheric dispersion factors (X/Qs) for Kewaunee Power Station have been revised and are listed in Table 1.3-3. Generated using the PAVAND code, the X/Qs are based upon a conservatively modeled ring with a 300-foot radius centered on Containment. This 300 foot bounding release ring, partially shown in Figure 3.1-1, encompasses all possible release points that exist within the station and is based upon the distance from the center of Containment to the farthest release point (i.e., Northeast Turbine Building corner). All actual release points are contained within this 300 foot bounding ring. The EAB and LPZ X/Q values were conservatively modeled using a ground-level release without credit for building wake effects.

Figure 2.2-2 in the KPS USAR shows the KPS EAB as an exclusion radius of 1,200 meters. The exclusion radius over land falls within the physical site boundary. For conservatism, the LPZ was calculated assuming the bounding shortest radius of 2 miles (3218.7 m). Utilizing the 300-foot (91.4 meters) bounding release ring described above, the shortest distance to the EAB (3,637 ft or 1,108.6 m) and the LPZ (10,260 ft or 3,127.3 m) for all directions (centered on the containment) was used to represent the bounding assumption for all possible release points. Modeled as a ground level release, the resulting EAB and LPZ X/Qs were determined by selecting the largest calculated value across all sixteen downwind directions and the overall site for each prescribed time period. The EAB (0-2 hour) X/Q is a single bounding value of 1.76E-04 sec/m3. The LPZ (0-8 hr, 8-24 hr, 1-4 day, and 4-30 day) X/Qs represent the highest calculated values for each time period across all directions. The maximum values occurred in the East-Northeast (ENE) direction for all except one time period, the (4-30 day) period, which occurred in the East (E) direction. Selecting the highest value within each time period across all directions and the overall site assures that the doses calculated for the LPZ are conservative.

Serial Number 11-025A Page 35 of 191 3.2 Design Basis Loss of Coolant Accident (LOCA) Reanalysis Amendment 190 (TAC No. MC9715, ADAMS Accession No. ML070430020) for Kewaunee Power Station, dated March 8, 2007 (Reference 14), represents the current licensing basis for the LOCA. This amendment incorporated TS changes to compensate for the higher control room emergency zone (CREZ) unfiltered in-leakage measured during the American Society for Testing and Materials (ASTM) E741 (tracer gas) leakage test conducted in December 2004 (Reference 20).

This section describes the methods employed and results obtained from the radiological reanalysis of the design basis LOCA. The analysis considers dose from several sources. They are:

Containment Leakage Plume,

Emergency Core Cooling System (ECCS) Component Leakage

Refueling Water Storage Tank Vent

Containment, Plume, and Filter Shine are negligible to control room occupants based on control room structure boundaries, penetration pathways and internal shield walls consisting of at least, or equivalent to, 18 inches of concrete; based on NUREG-0800, Section 6.4, Control Room Habitability System (Reference 21)

Containment purge isolates within 37 seconds following the LOCA and is an insignificant contributor to control room and offsite dose.

Doses are calculated at the Exclusion Area Boundary (EAB) for the worst-case two-hour period, at the Low Population Zone Boundary (LPZ), and in the KPS Control Room.

The methodology used to evaluate the doses resulting from a LOCA is consistent with RG 1.183 (Reference 1).

3.2.1 LOCA Scenario Description The design basis LOCA scenario for radiological calculations is initiated assuming a major rupture of the primary reactor coolant system piping. In order to yield radioactive

Serial Number 11-025A Page 36 of 191 releases of the magnitude specified in RG 1.183, it is also assumed that the ECCS does not provide adequate core cooling, such that significant core melting occurs. This general scenario does not represent any specific accident sequence, but is representative of a class of severe damage incidents that were evaluated in the development of the RG 1.183 source term characteristics. Such a scenario would be expected to require multiple failures of systems and equipment and lies beyond the severity of incidents evaluated for design basis transient analysis. Activity from the core is released to the containment, and from there released to the environment by means of containment leakage and leakage from the emergency core cooling system. For the containment leakage analysis, all activity released from the fuel is assumed to be in the containment atmosphere until removed by sprays, sedimentation, radioactive decay or leakage from the containment. For the ECCS leakage analysis, all iodine activity released from the fuel is assumed to be in the sump solution until removed by radioactive decay or leakage from the ECCS.

3.2.2 LOCA Source Term Definition RG 1.183 provides explicit description of the key AST characteristics recommended for use in design basis radiological analyses. The core radionuclide inventory used in this analysis was previously generated using the ORIGEN2 code for a Stretch Power Uprate (SPU) to 1772 megawatt thermal (MWt) and used in KPS Amendment No. 172, issued February 27, 2004 (Reference 11). Table 3.2-1 lists the RG 1.183 source term assumptions used in the LOCA analysis, which includes: the core inventory release fractions by radionuclide group, timing of release, and chemical form of the release into containment.

RG-1.183 divides the releases from the core into two phases:

1. The Fuel Gap Release Phase during the first 30 minutes and

2. The Early In-vessel Release Phase in the subsequent 1.3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months .

Serial Number 11-025A Page 37 of 191 Table 3.2-2 shows the fractions of the total core inventory of various isotope groups that are assumed released in each of the two phases of the LOCA analysis. Table 3.2-3 lists the isotopes and the associated curies at the end of a fuel cycle that was input to RADTRAD-NAI. The core inventory used in the LOCA analysis is the same source term used in Amendment No. 172, augmented with some additional core curies for Rb-88 and Cs-138 (Reference 11). Table 3.2-3 also provides the CEDE and EDE dose conversion factors for each of the isotopes. These dose conversion factors were taken from Federal Guidance Reports 11 and 12 (References 15 and 16, respectively).

Serial Number 11-025A Page 38 of 191 Table 3.2-1 Regulatory Guide 1.183 Source Terms Characteristic RG 1.183 Source Term Core Fractions Released To Containment Noble Gases 100%

Iodine 40%

Cesium 30%

Tellurium 5%

Barium 2%

Others - 0.02% to 0.25%

Timing of Release Released in Two Phases Over 1.8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Interval Iodine Chemical and Physical Form 4.85% Inorganic Vapor 0.15% Organic Vapor 95% Aerosol Solids Treated as an Aerosol Table 3.2-2 RG 1.183 Release Phases Core Release Fractionsa Isotope Group Gap Early In-Vessel Noble Gasesb 0.05 0.95 Halogens 0.05 0.35 Alkali Metals 0.05 0.25 Tellurium 0

0.05 Barium, Strontium 0

0.02 Noble Metals 0

0.0025 Cerium 0

0.0005 Lanthanides 0

0.0002 Duration (hours) 0.5 1.3

a. Release duration apply only to the Containment release. The ECCS leakage portion of the analysis conservatively assumes that the entire core release fraction is in the containment sump from the start of the LOCA.

b. Noble Gases are not scrubbed from the containment atmosphere and therefore are not found in either the sump or ECCS fluid.

Serial Number 11-025A Page 39 of 191 Table 3.2-3 Core Inventory and Dose Conversion Factors by Isotope (1782.6 MWt with 1.06 Multiplier*)

Isotope Isotope Group Curies EDE CEDE Sv-m3/Bq-sec Sv/Bq Kr-85 Noble gas 5.71E+05 1.190E-16 0.000E+00 Kr-85m Noble gas 1.39E+07 7.480E-15 0.000E+00 Kr-87 Noble gas 2.68E+07 4.120E-14 0.000E+00 Kr-88 Noble gas 3.77E+07 1.020E-13 0.000E+00 Xe-131m Noble gas 5.64E+05 3.890E-16 0.000E+00 Xe-133 Noble gas 9.98E+07 1.560E-15 0.000E+00 Xe-133m Noble gas 3.05E+06 1.370E-15 0.000E+00 Xe-135 Noble gas 2.77E+07 1.190E-14 0.000E+00 Xe-135m Noble gas 2.03E+07 2.040E-14 0.000E+00 Xe-138 Noble gas 8.65E+07 5.770E-14 0.000E+00 I-131 Halogen 5.04E+07 1.820E-14 8.890E-09 I-132 Halogen 7.33E+07 1.120E-13 1.030E-10 I-133 Halogen 1.04E+08 2.940E-14 1.580E-09 I-134 Halogen 1.14E+08 1.300E-13 3.550E-11 I-135 Halogen 9.73E+07 8.294E-14 3.320E-10 Rb-86 Alkali Metal 1.11E+05 4.810E-15 1.790E-09 Rb-88 Alkali Metal 3.77E+07 3.360E-14 2.260E-11 Cs-134 Alkali Metal 9.82E+06 7.570E-14 1.250E-08 Cs-136 Alkali Metal 2.80E+06 1.060E-13 1.980E-09 Cs-137 Alkali Metal 6.09E+06 2.725E-14 8.630E-09 Cs-138 Alkali Metal 8.65E+07 1.210E-13 2.740E-11 Sb-127 Tellurium 5.36E+06 3.330E-14 1.630E-09 Sb-129 Tellurium 1.62E+07 7.140E-14 1.740E-10 Te-127 Tellurium 5.31E+06 2.420E-16 8.600E-11 Te-127m Tellurium 6.90E+05 1.470E-16 5.810E-09 Te-129 Tellurium 1.59E+07 2.750E-15 2.090E-11 Te-129m Tellurium 2.35E+06 3.337E-15 6.484E-09

Serial Number 11-025A Page 40 of 191 Table 3.2-3 Core Inventory and Dose Conversion Factors by Isotope (1782.6 MWt with 1.06 Multiplier*)

Isotope Isotope Group Curies EDE CEDE Sv-m3/Bq-sec Sv/Bq Te-131 Tellurium 0.00E+00 2.040E-14 1.290E-10 Te-131m Tellurium 7.31E+06 7.463E-14 1.758E-09 Te-132 Tellurium 7.21E+07 1.030E-14 2.550E-09 Sr-89 Barium-Strontium 5.11E+07 7.730E-17 1.120E-08 Sr-90 Barium-Strontium 4.51E+06 7.530E-18 3.510E-07 Sr-91 Barium-Strontium 6.33E+07 4.924E-14 4.547E-10 Sr-92 Barium-Strontium 6.82E+07 6.790E-14 2.180E-10 Ba-139 Barium-Strontium 9.34E+07 2.170E-15 4.640E-11 Ba-140 Barium-Strontium 8.99E+07 8.580E-15 1.010E-09 Mo-99 Noble Metal 9.63E+07 7.280E-15 1.070E-09 Rh-105 Noble Metal 4.71E+07 3.720E-15 2.580E-10 Ru-103 Noble Metal 7.59E+07 2.251E-14 2.421E-09 Ru-105 Noble Metal 5.10E+07 3.810E-14 1.230E-10 Ru-106 Noble Metal 2.52E+07 1.040E-14 1.290E-07 Tc-99m Noble Metal 8.44E+07 5.890E-15 8.800E-12 Ce-141 Cerium 8.54E+07 3.430E-15 2.420E-09 Ce-143 Cerium 7.97E+07 1.290E-14 9.160E-10 Ce-144 Cerium 6.54E+07 2.773E-15 1.010E-07 Np-239 Cerium 1.01E+09 7.690E-15 6.780E-10 Pu-238 Cerium 1.90E+05 4.880E-18 7.790E-05 Pu-239 Cerium 1.93E+04 4.240E-18 8.330E-05 Pu-240 Cerium 2.67E+04 4.750E-18 8.330E-05 Pu-241 Cerium 6.24E+06 7.250E-20 1.340E-06 Am-241 Lanthanides 7.56E+03 8.180E-16 1.200E-04 Cm-242 Lanthanides 1.62E+06 5.690E-18 4.670E-06 Cm-244 Lanthanides 1.66E+05 4.910E-18 6.700E-05 La-140 Lanthanides 9.76E+07 1.170E-13 1.310E-09

Serial Number 11-025A Page 41 of 191 Table 3.2-3 Core Inventory and Dose Conversion Factors by Isotope (1782.6 MWt with 1.06 Multiplier*)

Isotope Isotope Group Curies EDE CEDE Sv-m3/Bq-sec Sv/Bq La-141 Lanthanides 8.53E+07 2.390E-15 1.570E-10 La-142 Lanthanides 8.26E+07 1.440E-13 6.840E-11 Nb-95 Lanthanides 8.77E+07 3.740E-14 1.570E-09 Nd-147 Lanthanides 3.40E+07 6.190E-15 1.850E-09 Pr-143 Lanthanides 7.70E+07 2.100E-17 2.190E-09 Y-90 Lanthanides 4.68E+06 1.900E-16 2.280E-09 Y-91 Lanthanides 6.55E+07 2.600E-16 1.320E-08 Y-92 Lanthanides 6.85E+07 1.300E-14 2.110E-10 Y-93 Lanthanides 7.87E+07 4.800E-15 5.820E-10 Zr-95 Lanthanides 8.71E+07 3.600E-14 6.390E-09 Zr-97 Lanthanides 8.62E+07 4.432E-14 1.171E-09

increased by 6% to account for variations in core parameters: 493.6 +/- 10% EFPD, average enrichment of 4.5 w/o +/- 10%, and core mass of 49.1 MTU +/- 10%.

Although Te-131 was not included in the initial core inventory, it was included in the analysis as a significant decay product.

3.2.3 LOCA Atmospheric Dispersion Factors 3.2.3.1 LOCA Control Room /Qs As described in Section 3.1, the onsite atmospheric dispersion factors were calculated using the ARCON96 code (Reference 5) and guidance from Regulatory Guide 1.194 (Reference 6). The LOCA Control Room X/Qs listed in Table 1.3-4 were calculated for the following KPS source points:

Reactor Building Exhaust Stack

Shield Building

Auxiliary Building Exhaust Stack

Serial Number 11-025A Page 42 of 191 3.2.3.2 LOCA Offsite (EAB & LPZ) /Qs As described in Section 3.1, the Exclusion Area Boundary (EAB) and Low Population Zone (LPZ) atmospheric dispersion factors were revised and are listed in Table 1.3-3.

These offsite atmospheric dispersion factors were generated using the PAVAND code (Reference 7) and guidance from Regulatory Guide 1.145 (Reference 8).

3.2.4 LOCA Containment Airborne Activity 3.2.4.1 Containment Sprays The current licensing basis for the LOCA uses containment sprays to remove elemental and particulate iodine from the containment atmosphere. The use of containment sprays and methods to determine elemental and particulate iodine removal rates were approved in KPS Amendment No. 166, issued March 17, 2003 (Reference 10).

One train of the containment spray system is assumed to operate following the LOCA.

Injection spray is credited with no delay in startup. Earlier spray actuation is conservative since it results in earlier spray termination. There is no benefit from earlier spray actuation since there is little activity in the containment at the time the spray starts. When the RWST drains to a predetermined setpoint level, the operators switch to recirculation of sump liquid. Switchover to recirculation spray is not credited in the analysis and all spray is assumed to be terminated when the RWST drains down. The analysis conservatively assumes that the sprays are terminated 0.91 hours0.00105 days 0.0253 hours 1.50463e-4 weeks 3.46255e-5 months after the start of the event. New spray removal rates were determined based on the revised assumptions.

KPS containment spray design consists of four spray ring headers. The elemental and particulate iodine removal rates due to sprays are listed in Table 3.2-5. These spray removal rates are used until the RWST is secured at 0.91 hours0.00105 days 0.0253 hours 1.50463e-4 weeks 3.46255e-5 months . At that time, further iodine removal is ignored due to sprays even though the recirculation spray system remains operating. An elemental iodine DF of 200 and particulate iodine DF of 50 are

Serial Number 11-025A Page 43 of 191 not achieved during the period that sprays are assumed operating. Therefore, the elemental and particulate iodine removal rates remain constant during this period.

3.2.4.1.1 Containment Spray Removal of Elemental Iodine NUREG-0800, Section 6.5.2, Rev. 2 (Reference 22) identifies a methodology previously used and approved for the determination of spray removal of elemental iodine. The removal rate constant is determined by:

= (6 Kg T F) / V D where;

= elemental iodine removal coefficient, Kg

= Gas phase mass transfer coefficient T

= Time of fall of the spray drops F

= Volume flow rate of sprays V

= Containment sprayed volume D

= Mass-mean diameter of the spray drops The spray parameter values are listed in Table 3.2-4.

These parameters and the appropriate conversion factors were used to calculate the elemental spray removal coefficient. The calculated value of 15 hr-1 is modeled for removal of elemental iodine from the containment atmosphere. The maximum DF of 200 is not achieved prior to assumed spray termination at 0.91 hours0.00105 days 0.0253 hours 1.50463e-4 weeks 3.46255e-5 months .

3.2.4.1.2 Containment Spray Removal of Particulates The particulate removal coefficient was calculated using a Regulatory Guide 1.183 prescribed method from NUREG/CR-5966, A Simplified Model of Aerosol Removal by Containment Sprays (Reference 23). Inputs to the methodology include the fall height

Serial Number 11-025A Page 44 of 191 H of the water droplets in meters and the spray water flux Q in (cm3 of H2O)/(cm2sec). Values for the fall height and spray water flux are given in Table 3.2-4.

The current particulate removal coefficient is calculated using the model described in NUREG-0800, Section 6.5.2, Revision 2 (Reference 22). Both the NUREG/CR-5966 and SRP 6.5.2 methods are deemed acceptable per RG 1.183. DEK has elected to change to the NUREG/CR-5966 method for KPS based solely on commonality to methods used at other Dominion facilities. A comparison of resulting particulate removal constants from both methods was made to determine if this change in methodology provides a benefit. The NUREG/CR-5966 method produces a smaller, more conservative coefficient that is used in the revised LOCA analysis.

NUREG/CR-5966 [Page 173] presents the following equations for aerosol (i.e., aerosol treated as particulate in SRP methodology) removal rate at the 10th percentile level:

8945

.0 8945

.0 10 9.0 2

2 2

2 9.0 9.0 9.0 1

log 00201

.0 1108

.0 6

E 555

.3 3

E 9821

.6 7

E 327

.7 ln 94362

.0 5750

.5 ln

f f

m m

m Q

H Q

QH Q

f f

f

where is the removal rate, mf is the mass fraction remaining in the containment, H is the spray drop height, and Q is the spray water flux, calculated by dividing the spray flow rate (F) by the wetted cross-sectional area of the sprayed portion of the containment. The wetted cross-sectional area is determined by multiplying the containment cross-sectional area (A) by the sprayed fraction (SF). The inner radius of containment is 52.5 ft, yielding a cross-sectional area of 8.659E3 ft2. The first equation above is used to calculate the removal rate corresponding to a mass fraction of 0.9. Substituting this value into the second equation yields the removal for a given value of mass fraction. Since the removal rate is dependent on drop height and spray rate, the smallest (most conservative) value for each is used to calculate the lowest removal rate that will be applied over the entire time when sprays are credited.

Serial Number 11-025A Page 45 of 191 Spray flux is derived as follows:

Q = (F gpm) (6.791E-2) / (A ft2 x SF)

[conversion]

Table 3.2-4 presents the spray fall height of 116.5 ft which was derived by subtracting the 649'-6" elevation of the refueling floor from the weighted-average spray header elevation of 766-0 (Reference 24), assuming one train of spray pumps available.

NUREG/CR-5966 [Page 170] recommends that for a volume with continuing source, the spray removal constant associated with a mass fraction of 0.9 be used until the time-dependent source terminates. Hence, the mass fraction should be assumed to remain at 0.9 from the start of the sprays until the end of the early in-vessel release phase at 1.8 hr. The resulting spray removal coefficient for mf = 0.9 is 2.855 hr-1, rounded down conservatively to 2.8 hr-1.

The spray removal coefficient of 2.8 hr-1 was used over the period that sprays are credited. The airborne inventory does not drop to 2 percent of the total particulate iodine released to the containment (i.e., a DF of 50) before spray termination at 0.91 hours0.00105 days 0.0253 hours 1.50463e-4 weeks 3.46255e-5 months .

Table 3.2-5 lists the aerosol and elemental iodine removal coefficients determined for KPS.

3.2.4.2 Natural Deposition A reduction in airborne radioactivity in the containment by natural deposition within containment is credited. The model used is described in NUREG/CR-6189, A Simplified Model of Aerosol Removal by Natural Processes in Reactor Containments, (Reference 25) and is incorporated into the RADTRAD-NAI computer code. This model is called the Powers model, set for the 10th percentile.

Serial Number 11-025A Page 46 of 191 3.2.5 LOCA Analysis Assumptions & Key Parameter Values 3.2.5.1 Method of Analysis The RADTRAD-NAI code (Reference 3) is used to calculate the radiological consequences from airborne releases to the EAB, LPZ, and Control Room resulting from a LOCA at Kewaunee Power Station (KPS).

RADTRAD can model a variety of processes that can attenuate and/or transport radionuclides. It can model sprays, filtered flow, and natural deposition that reduce the quantity of radionuclides suspended in the containment or other compartments. The RADTRAD models used in this calculation include the following pathways:

Activity from the failed fuel enters the containment and is released to the atmosphere through containment leakage. All nuclides are released through this pathway. This pathway is not filtered.

Containment air enters the shield building. A portion of the shield building air volume is discharged to the environment as necessary to maintain a negative pressure. Releases from the shield building to the environment are filtered.

Negative pressure in the shield building is established within 10 minutes of the accident. During the first 10 minute interval, no credit is taken for filtering the shield building exhaust.

Containment air enters the auxiliary building Special Ventilation (SV) zone to the environment. Releases from the SV zone to the environment are filtered.

Activity in the sump leaks out of containment via the ECCS system and is released to the auxiliary building SV zone and then to the environment. Only iodine is released through this pathway. This pathway is filtered.

Activity in the ECCS back-leaks to the RWST. The RWST vents into the auxiliary building and is captured by SV before exhausting to the environment. This pathway is filtered.

The revised LOCA analysis contains some changes to the plant specific assumptions and methods. These changes include:

Serial Number 11-025A Page 47 of 191

Conservative increase in core radionuclide curie inventory by applying a 1.06 multiplier to account for fuel management variations

Conservative recalculation of spray removal coefficients based on a reduced spray droplet fall height

Replacement of a sedimentation removal coefficient of 0.1 hr-1 with the Powers model built into RADTRAD

Recalculation of offsite and control room X/Q dispersion factors

Conservative increase in assumed iodine evolution rate from ECCS leakage to 10% for the entire 30-day duration of leakage. Current analysis of record assumes 10% evolution for 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months post accident, then 1% thereafter.

Replacement of the assumed 1% iodine evolution rate from RWST back-leakage to a conservative DF=100.

The combined effect of these changes result in changes to the EAB, LPZ, and control room doses due to a KPS design basis LOCA. In all cases, the doses fall within required limits.

3.2.5.2 Basic Data & Assumptions for LOCA Changes have been made to the AST LOCA. Tables 3.2-4 and 3.2-5 provide a complete list of inputs and assumptions used to reanalyze the KPS LOCA.

Serial Number 11-025A Page 48 of 191 Table 3.2-4 Spray Removal Calculation Parameters Parameter or Assumption CLB Value Proposed Value Reason for Change Elemental Iodine Removal Coefficient Kg Gas phase mass transfer coefficient 3 m/min No change T

Spray drop fall time 13 seconds 9 seconds Shorter fall height F

Volume flow rate of sprays 1148 gpm = 9,208 ft3/hr No change V

Containment sprayed volume 1.32E6 ft3 No change D

Mass-mean diameter of the spray drops 1210 m = 3.97E-3 ft No change Particulate Iodine Removal Coefficient H

Fall Height of droplets 150 ft 116.5 ft Average spray header height to the refueling floor.

Q Spray Water Flux Not Applicable Derived value = F/(A

SF) 9.003E-3 cm3 H2O/cm2-s New method employed:

NUREG/CR-5966 A

Cross sectional area of containment 8.659E3 ft2 No change SF Sprayed fraction in containment 1.0 No change

Serial Number 11-025A Page 49 of 191 Table 3.2-5 Basic Data and Assumptions for LOCA Parameter or Assumption CLB Value Proposed Value Reason for Change Source Term Core Power (MWt) 1782.6 (Licensed power of 1772 MWt with 0.6%

uncertainty)

No Change Core Inventory (curies)

Licensed Uprated Core based on 1782.6 MWt multiplied by 1.03 to account for fuel management variations Licensed Uprated Core based on 1782.6 MWt multiplied by 1.06 to account for fuel management variations Conservative assumption Dose Conversion Factors CEDE Whole Body Values are from Table 2.1 of Federal Guidance Report (FGR) 11 ICRP 30 [Westinghouse TITAN5 code]

No Change Table III.1 of FGR 12 Per RG 1.183 Core Release Fraction, Gap Release Fractions and Release Timing Values from Table 2 and 4 of RG 1.183 No Change Initial Iodine Species in Containment (%)

Elemental Methyl (organic)

Particulate (aerosol) 4.85 0.15 95 No Change

Serial Number 11-025A Page 50 of 191 Table 3.2-5 Basic Data and Assumptions for LOCA Parameter or Assumption CLB Value Proposed Value Reason for Change Containment Containment Leak Rate (wt%/day) 0-24 hours

>24 hours 0.2 0.1 No Change Containment Leak Path Fractions 0-10 minutes Through Shield Bldg Through Aux Bldg SV Direct to Environment 10 minutes - 30 days Through Shield Bldg Through Aux Bldg SV Direct to Environment 0.0 0.10 0.90 0.89 0.10 0.01 No Change Shield Building Drawdown Time: (Tech Specs) 10 minutes No Change Containment Volume (ft3) 1.32E6 No Change Containment Purge Release Prior to Containment Isolation Not Analyzed Negligible KPS is a licensed leak-before-break LBB plant (Reference 9). Per RG 1.183, the onset of gap release can be credited with a 10 minute delay for LBB.

Containment purge isolation

Serial Number 11-025A Page 51 of 191 Table 3.2-5 Basic Data and Assumptions for LOCA Parameter or Assumption CLB Value Proposed Value Reason for Change occurs within 37 seconds.

Therefore, dose contribution from only TS RCS inventory is insignificant.

Containment Sump and Sprays Iodine Chemical Form in the Sump (%)

100% Elemental 97% Elemental 3% Organic Per RG 1.183 Containment Sump pH:

at least 7 No Change Containment Sump Volume (gal) 315,000 311,000 Subtracted tank volume measurement uncertainties Containment Spray Coverage

(%)

100 No Change Containment Spray Duration (hr) 0.917 0.91 Based on revised RWST low level signal based on minimum drain down time Containment Spray Recirculation No credited No Change Containment spray Removal Coefficient (hr-1)

Elemental Particulate 20 4.5 15 2.8 Reduction in coefficient values is primarily a result of reduced droplet fall height

Serial Number 11-025A Page 52 of 191 Table 3.2-5 Basic Data and Assumptions for LOCA Parameter or Assumption CLB Value Proposed Value Reason for Change Natural deposition (hr-1) 0.1 Powers Model set at the 10th percentile Per RG 1.183 Shield Building Shield Building Annulus Volume (ft3) 3.74E+05 No Change Shield Building Participation Fraction 0.5 No Change Shield Building Ventilation and Recirculation Iodine Filter Efficiency (%)

Elemental Methyl (organic)

Particulate (aerosol) 95 (includes safety factor of 2) 95 (includes safety factor of 2) 99 No Change Shield Building Air Flow to Environment (cfm) 0-10 min 10-30 min

>30 min 0

6600 3100 No Change Shield Building Recirculation Flow (cfm) 0-30 min

>30 min 0

2300 No Change

Serial Number 11-025A Page 53 of 191 Table 3.2-5 Basic Data and Assumptions for LOCA Parameter or Assumption CLB Value Proposed Value Reason for Change Auxiliary Building Participation with Auxiliary Building Volume or Hold-up None No Change Auxiliary Building Special Ventilation Iodine Filter Efficiency (%)

Elemental Methyl (organic)

Particulate (aerosol) 95 (includes safety factor of 2) 95 (includes safety factor of 2) 99 No Change ECCS ECCS Leak Rate to Auxiliary Building (gal/hr) 12 (twice the leakage limit)

No Change ECCS Iodine Airborne Evolution (%)

0-3 hour

>3 hour 10 1

10 10 Conservative change using RG 1.183 guidance Plate-out in Aux Bldg (%)

50 No Change

Serial Number 11-025A Page 54 of 191 Table 3.2-5 Basic Data and Assumptions for LOCA Parameter or Assumption CLB Value Proposed Value Reason for Change Start of Recirculation (hr)

Conservatively assumes leakage starts at 0 hours0 days 0 hours 0 weeks 0 months No Change RWST RHR Back-Leakage to RWST (gpm) 0-24 hour 1-30 day 3

1.5 No Change Start of Back-Leakage (hr)

Conservatively assumes leakage starts at 0 hours0 days 0 hours 0 weeks 0 months No Change RWST Iodine Airborne Evolution 1%

DF=100 Change in methodology consistent with and approved at other Dominion facilities (e.g., Millstone Unit 3 and North Anna). DF of 100 is conservative compared to a calculated DF greater than 300 for the KPS RWST.

Plate-out in Aux Bldg (%)

50 No Change

Serial Number 11-025A Page 55 of 191 Table 3.2-5 Basic Data and Assumptions for LOCA Parameter or Assumption CLB Value Proposed Value Reason for Change Core Total Iodine Mass (kg)

Not Used 14.23 Partition Coefficient in RWST is dependent upon total iodine. Value is conservatively based on ORIGEN results; Mass ratio of Total Iodine to Iodine-131 equals 35.

Maximum Sump Total Iodine Concentration (mg/liter)

Not Used 4.83 Based on sump volume and 40% of the core total iodine.

RWST total iodine concentration is conservatively maximized to minimize iodine partition coefficient.

Maximum RWST Total Iodine Concentration (mg/liter)

Not Used 3.05 Maximum RWST total iodine concentration is achieved at 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months . Maximum concentration results in lowest partition coefficient (PC), from Ref. 26: 3.05 mg/l results in a PC=581.

RWST Tank Volume (gal)

Not Used 272,500 [3.64E4 ft3 ]

RWST DF is a function of tank liquid and air volumes Applied RWST DF Not Used 100 Conservative value

Serial Number 11-025A Page 56 of 191 Table 3.2-5 Basic Data and Assumptions for LOCA Parameter or Assumption CLB Value Proposed Value Reason for Change Time Dependent RWST Liquid Volume from Back-Leakage Not Used Time (hrs) 0 3

6 12 24 48 96 200 400 720 Liquid (ft3) 5253 5325 5397 5542 5830 6119 6697 7948 10354 14204 Calculated Values. New partition coefficient method requires calculation of RWST volumes and concentrations.

Calculated Time Dependent RWST DF Not Used Time (hrs) 0 3

6 12 24 48 96 200 400 720 DF 841 846 818 758 626 525 413 336 338 372 Calculated DF values consider time dependent RWST liquid and air volume and increasing iodine concentrations in the RWST. Over the entire 30 day accident, calculated DF values are greater than a factor of three above the applied DF value used in the RWST release analysis.

EAB X/Q (sec/m3) 0 - 2 hr 2.232E-04 1.76E-04 New PAVAND X/Q values (see Table 1.3-3 and Section 3.1.2)

Serial Number 11-025A Page 57 of 191 Table 3.2-5 Basic Data and Assumptions for LOCA Parameter or Assumption CLB Value Proposed Value Reason for Change LPZ X/Q (sec/m3)

Period LPZ 0 - 2 hr 3.977E-05 2 - 24 hr 4.100E-06 1 - 2 day 2.427E-06 2 - 30 day 4.473E-07 Period LPZ 0 - 8 hr 3.36E-05 8 - 24 hr 2.37E-05 1 - 4 day 1.12E-05 4 - 30 day 3.94E-06 New PAVAND X/Q values (see Table 1.3-3 and Section 3.1.2)

Control Room Control Room Volume (ft3) 127,600 No Change Normal Ventilation Unfiltered Makeup Air Flow (scfm) 2,500 (nominal) 2,750 Maximum flow considering

+/- 10% uncertainty Filtered Recirculation Air Flow (scfm) 2,250 No Change CRPARS Filter Efficiency (%)

Elemental Organic Particulate 90 (includes safety factor of 2) 90 (includes safety factor of 2) 99 No Change

Serial Number 11-025A Page 58 of 191 Table 3.2-5 Basic Data and Assumptions for LOCA Parameter or Assumption CLB Value Proposed Value Reason for Change Control Room Isolation (sec) 120 0

Control room isolation damper takes 10 seconds to close upon receipt of SI signal which is generated within seconds of the LOCA.

Due to AST release delay of 30 seconds per RG 1.183, the control room will be isolated prior to any radioactive release.

CRPARS Start (sec) 120 133 Based on 10 second delay to switchover from normal ventilation to emergency operation, 63 second delay in diesel loading of CRPARS, and 60 seconds to open recirculation dampers Control Room Unfiltered Inleakage (cfm) 800 No Change Maximum (ASTM) E741 tracer gas test in Dec 2004 was 447+/-51 cfm (Ref. 20)

Serial Number 11-025A Page 59 of 191 Table 3.2-5 Basic Data and Assumptions for LOCA Parameter or Assumption CLB Value Proposed Value Reason for Change Release point(s)

Containment / Shield Bldg Rx Building Stack Exhaust Aux Building Stack Exhaust No Change Control Room X/Q (sec/m3)

Containment / Shield Bldg Rx Bldg Stack Exhaust Aux Bldg Stack Exhaust for all releases 0 - 8 hrs 2.93E-03 8 - 24 hrs 1.73E-03 24 - 96 hrs 6.74E-04 96 - 720 hrs 1.93E-04 0 - 2 hr 1.74E-03 3.97E-03 2.90E-03 New ARCON96 control room X/Q estimates (Table 1.3-4)

Prior to plume arrival, normal control room intake will isolate. X/Q values represent the worst case unfiltered inleakage location.

For period values out to 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months , see Table 1.3-4

Serial Number 11-025A Page 60 of 191 3.2.5.3 LOCA Containment Leakage Model Containment leakage consists of filtered and bypass leakage. The total containment leak rate (La) is 0.2% per day (weight %/day) for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Thereafter, the leak rate is one half or 0.1% per day for the remaining accident duration, out to 30 days.

For the first 10 minutes following the LOCA, the Shield Building is ignored while it is pumping down to vacuum conditions. Releases from containment are split, 10% being released from the filtered Auxiliary Building Ventilation (ABV) exhaust stack and the remaining 90% being released at ground level directly to the environment. The 10 minute interval is conservative because a measureable vacuum is developed in the shield building within 4 minutes of the fan startup.

For the first 30 minutes, the recirculation of Shield Building annulus air is ignored. After vacuum conditions are achieved at 10 minutes, releases are assumed to begin out the filtered Shield Building Ventilation (SBV) exhaust stack at a conservatively high rate of 6600 cfm (highest starting drawdown rate prior to vacuum conditions). The percentage of bypass leakage assumed to escape directly to the environment is reduced to 1% of La with 10% continuing out the filtered ABV and the remaining 89% through the filtered SBV.

After 30 minutes post LOCA, Shield Building recirculation is credited. The split of containment releases remain 10%, 89% and 1% between the ABV, SBV and direct to the environment. The Shield Building requires an exhaust rate less than 2000 cfm to maintain vacuum conditions once achieved at 30 minutes, the analysis conservatively assumes 3100 cfm.

The collection, processing, and release of containment leakage vary depending on the location of the leak. Ventilation characteristics and release paths are different for the Shield Building and Auxiliary Building. KPS Technical Specification 5.5.14 leakage acceptance criteria provide the basis for release assumptions for containment leakage.

Serial Number 11-025A Page 61 of 191 Figure 3.2-1 displays the assumptions, inputs and pathways used in RADTRAD to model KPS containment airborne releases from a design basis LOCA.

Figure 3.2-1 RADTRAD Model for Containment Airborne Releases 3.2.5.4 Model of ECCS Leakage The Emergency Core Cooling System (ECCS) fluid consists of the contaminated water in the sump of the containment. This water contains 40% of the core inventory of iodine, 5% released to the sump water from the gap release phase and 35% released to the sump water from the early in-vessel phase. During a LOCA, the highly radioactive fluid is pumped from the containment sump to the recirculation spray headers and sprayed back into the containment sump. Also, following a design basis LOCA, valve

Serial Number 11-025A Page 62 of 191 realignment occurs to switch the suction water source for the ECCS pumps from RWST to the containment sump.

ECCS leakage develops when ESF (engineered safeguards feature) systems circulate sump water outside containment and leaks develop through packing glands, pump shaft seals and flanged connections. Station procedures specify a limit of 6 gallons per hour for total allowed ECCS leakage. In accordance with RG 1.183, the ECCS analysis makes use of two times the sum of the simultaneous leakage from all components in the ESF recirculation systems, or 12 gallons per hour for ECCS leakage. The leakage of recirculating sump fluids commences at 0.91 hours0.00105 days 0.0253 hours 1.50463e-4 weeks 3.46255e-5 months , which is the earliest time of recirculation. The analysis conservatively assumes leakage starts at 0 hours0 days 0 hours 0 weeks 0 months .

The temperature of the containment sump is conservatively assumed to reach a maximum of 293 degrees F (saturation conditions). At this maximum temperature, a flash fraction of less than 0.1 is calculated. Current analysis assumptions reduce the flash fraction to 0.01 after 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months when sump temperature drops below 212°F.

However, per the guidance of RG 1.183, a conservative flash fraction of 0.1 is used for the ECCS leakage during the entire event for all sump temperatures. The water volume of the sump is 311,000 gallons and is assumed to remain constant.

Per KPS Licensing Basis, a 50% plate-out of iodine evolved from flashing ECCS fluid is credited on surfaces in the large Auxiliary Building volume. The xenon progeny from the iodine that plates out is included in the dose analyses. Dilution and decay within the Auxiliary Building volume are not credited.

Figure 3.2-2 displays the assumptions, inputs and pathways used in RADTRAD to model KPS ECCS Leakage from a design basis LOCA.

Serial Number 11-025A Page 63 of 191 Figure 3.2-2 RADTRAD Model for ECCS Leakage into the Auxiliary Building 3.2.5.5 Model of ECCS Back Leakage to Refueling Water Storage Tank Following a design basis LOCA, valve realignment occurs to switch the suction water source for the ECCS pumps from the Refueling Water Storage Tank (RWST) to the containment sump. This switch occurs when the level in the RWST reaches a defined setpoint and is modeled in RADTRAD-NAI as occurring at 0.91 hours0.00105 days 0.0253 hours 1.50463e-4 weeks 3.46255e-5 months following the initiation of the LOCA. In this configuration, MOVs and check valves in the normal suction line from the RWST and MOVs in the recirculation line provide isolation between this contaminated flow stream and the RWST. This RADTRAD-NAI analysis of the LOCA models leakage of ECCS fluid through these valves back into the RWST with subsequent leakage of the evolved iodine through the vent of the KPS RWST into the Auxiliary Building.

The RADTRAD-NAI source term used to model the ECCS leakage into the RWST contains only the iodine isotopes. This is because iodine is the only element in the

Serial Number 11-025A Page 64 of 191 containment sump water which was modeled as coming out of solution and becoming airborne. Forty percent of the core inventory of iodine isotopes were conservatively modeled as being instantaneously transported from the core to the containment sump.

This iodine is modeled to be 97% in the elemental chemical form and 3% in the organic chemical form in accordance with RG-1.183.

The following two flowcharts shown in Figures 3.2-3 and 3.2-4 demonstrate the compartments and pathways used in RADTRAD to calculate the doses resulting from containment sump back-leakage into the RWST. Two separate models were used.

The first models the RWST liquid space as a variable volume and was used to calculate doses due to the release of iodines. This model reduces the flow rate from the RWST to the environment to reflect the iodine partition coefficient in the RWST. Additionally, iodines released from the RWST vent into the Special Ventilation (SV) zone within the Auxiliary Building and get filtered prior to exhaust from the Auxiliary Building Stack.

This model under-predicts the release of xenon isotopes produced from the decay of radioiodines in the RWST. The doses resulting from xenon are calculated using a second RWST release model.

Since the release of iodine is accounted for in the first model, a second model captures all iodine released from the sump in a 100% efficient iodine filter. All xenon resulting from iodine decay is released from the RWST out of the Auxiliary Building Stack. The combined doses resulting from the two RWST release models will conservatively predict the doses resulting from the iodine isotopes and their progeny.

Per KPS Licensing Basis, a 50% plate-out of iodine evolved from the RWST is credited on surfaces in the large Auxiliary Building volume. The xenon progeny from the iodine that plates out is included in the dose analyses. Dilution and decay within the Auxiliary Building volume is not credited.

The release scenario considers containment sump liquid leaking into the lines leading to the RWST and ignores any time delay that physically would occur as a result of

Serial Number 11-025A Page 65 of 191 contaminated sump fluid pushing clean fluid residing in the lines back to the RWST.

Back-leakage to the RWST is conservatively assumed to start at 0 hours0 days 0 hours 0 weeks 0 months . The release of radioactivity is a result of partitioning between the contaminated fluid within the RWST and the air sitting above the fluid within the tank. To maintain an equal pressure within the tank, the amount of air released equals the volume of sump fluid that leaks into the tank. Over the 1 to 30 day period following the accident when 1.5 gallons per minute of sump fluid is assumed to leak into the RWST, the release of air is 0.201 ft3/min.

A critical parameter in the radiological-impact analysis is the definition of a proper Partition Coefficient (PC) for the iodines in the RWST water. The PC applicable to the iodines in the RWST water was based on information in A. K. Postma, L. F. Coleman and R. K. Hilliard (Reference 26), Iodine Removal from Containment Atmospheres by Boric Acid Spray, Report No. BNP-100, Battelle Memorial Institute, Pacific Northwest Laboratories (PNL), Richland, WA 99352 (7/1970). Use of BNP-100 is discussed in SRP 6.5.2 (Reference 22). For this application, the RWST is assumed to behave like a closed system for the establishment of equilibrium conditions between the water and air.

This is the same method Dominion has employed at Millstone Unit 3 and North Anna submittals (References 33 and 34) for RWST releases due to sump back-leakage.

The critical factor in determining the magnitude of the PC is the total iodine concentration in the RWST water (on a mass basis, including stable iodine). Based on ORIGENS runs performed for Millstone Unit 3 and North Anna, which are both higher power PWRs (respectively 3650 MWt and 2940 MWt, compared to KPS at 1772 MWt),

the mass of Total Iodine to Iodine-131 is 32.6 and 30.7, respectively. A conservative mass ratio of 35 was used to approximate the Total Iodine mass of 14.23 kg in the KPS core. Assuming 40% of the core iodine is released and conservatively contained in the sump, a sump iodine mass of 5.69 kg is assumed to maximize sump iodine concentration at 4.83 mg/liter. The maximum RWST iodine concentration of 3.05 mg/liter occurred at 30 days. The PNL report (Reference 26), shows how higher iodine

Serial Number 11-025A Page 66 of 191 concentration results in lower partitioning. The PC predicted at 3.05 mg/liter is approximately 581.

The iodine decontamination factors associated with the releases from the RWST were calculated using the relationship (from Reference 22, Standard Review Plan, Section 6.5.2):

DF = 1 + (Vliq / Vair) PC where Vliq and Vair are the liquid and air volumes between which the partitioning takes place. Using the smallest ratio of Vliq to Vair at the onset of back-leakage will predict the smallest predicted DF. Used in conjunction with the lowest PC of 581, applicable for the worst case tank concentration, a DF greater than 100 is obtained. A DF of 100 is typically employed in many applications in the nuclear power industry and has previously been demonstrated as being conservative for RWST release applications for Millstone Unit 3 and North Anna.

Actual time dependent values of DF calculated for the RWST over the 30-day accident are shown in Table 3.2-6. These DFs exist at various times using calculated RWST iodine concentrations and ratios of Vliq to Vair to show that an assumed DF of 100 provides at least a factor of three conservatism over the entire accident. The modeling of RWST releases over the entire 30-day accident period credits only a PC value of 155, applicable to final RWST iodine concentrations and volumes that would yield a DF equal to 100. Modeling the entire release period with a constant partition coefficient provides additional conservatism beyond the factor of three already discussed. As Table 3.2-6 shows, actual partitioning in the tank can be more than an order of magnitude greater.

Using the methodology employed at Millstone Unit 3 and North Anna to model RWST back-leakage releases, the application of the model incorporates numerous

Serial Number 11-025A Page 67 of 191 conservatisms to assure predicted RWST radioactive releases are adequately bounding for the KPS LOCA analysis.

Table 3.2-6 RWST Time Dependent DF Values Parameter Value DF Determination (RWST Liquid volume, RWST Air volume, PC and DF)

Time Liquid Air PC*

DF**

(hrs)

(ft3)

(ft3) 0 5253 31180 4982 841 3

5325 31108 4936 846 6

5397 31036 4699 818 12 5542 30892 4218 758 24 5830 30603 3279 626 48 6119 30314 2597 525 96 6697 29737 1831 413 200 7948 28485 1199 336 400 10354 26079 849 338 720 14204 22229 581 372

Partition Coefficients taken from PNL Report [Reference 26] based on RWST total iodine concentration

- SRP 6.52 [Reference 22], DF = 1 + (Vliq / Vair) PC

Serial Number 11-025A Page 68 of 191 Figure 3.2-3 RADTRAD Model for Iodine Back-Leakage into the RWST

Serial Number 11-025A Page 69 of 191 Figure 3.2-4 RADTRAD Model for Noble Gas Leakage from RWST 3.2.5.6 KPS Control Room The control room volume is 127,600 ft3. The LOCA causes a Safety Injection (SI) signal, which also isolates the control room (per current Licensing Basis). The control room is isolated within 10 seconds after the SI signal. Based on RG 1.183, the onset of the gap release does not start until 30 seconds post-LOCA. Therefore, the control room will be isolated prior to the arrival of the radioactive release.

Serial Number 11-025A Page 70 of 191 Control room parameters are provided in Tables 1.3-1, 1.3-5, and 3.2-5. These parameters include the normal operation flow rates, the emergency operation flow rates, control room volume, filter efficiencies and control room operator breathing rates. In the analyses presented in this report, the control room is modeled as a discrete volume.

The Table 1.3-4 atmospheric dispersion factors are calculated to determine the activity available for intake into the control room from releases. The inflow to the control room and the control room recirculation flow are used to calculate the activity introduced to the control room and cleanup of activity from that flow. The control room filter efficiencies are conservatively assumed at 90% for both elemental and organic and 99% for aerosol iodine.

The CR ventilation system provides a large percentage of recirculated air. Process radiation monitor channel R-23 monitors control room ventilation air for radiation. If a high radiation condition exists, the monitor initiates closure of the outside air intake and starts the CR post accident recirculation system (CRPARS). KPS control room isolation and start of CRPARS also occurs on either a Safety Injection or Steam Exclusion signal.

In addition, local CR area radiation monitor channel R-1 monitors CR air for radiation and alarms when it reaches the CR area radiation monitor setpoint. No credit is taken for the alarms or automatic actions from R-23 and R-1 in the design basis LOCA.

The post LOCA dose consequences to the KPS control room are due to the following sources:

Containment leakage

ESF leakage

RWST backflow External shine sources are negligible to the overall control room dose consequences due to control room structure boundaries, penetration pathways and internal shield walls consisting of at least, or equivalent to, 18 inches of concrete (Reference 21).

Serial Number 11-025A Page 71 of 191 3.2.6 LOCA Results Table 3.2-7 lists TEDE to the EAB and LPZ from a LOCA at KPS. The dose to the EAB and LPZ is less than the 25 rem TEDE limit stated in 10 CFR 50.67 and Regulatory Guide 1.183. The EAB dose represents the worst 2-hour dose for each release pathway.

The dose to the KPS control room is less than the 5 rem TEDE limit specified in 10 CFR 50.67 and Regulatory Guide 1.183.

Table 3.2-7 Dose summary for a Kewaunee LOCA Location TEDE (rem)

Limits (rem)

EAB 0.5 25 LPZ 0.5 25 Control Room 4.1 5

Serial Number 11-025A Page 72 of 191 3.3 Fuel Handling Accident (FHA)

This section describes the methods employed and results of the Fuel Handling Accident (FHA) design basis radiological analysis. The analysis includes doses associated with release of gap activity from a fuel assembly either inside containment or in the Spent Fuel Pool. Doses were calculated at the Exclusion Area Boundary (EAB), at the Low Population Zone (LPZ) boundary, and in the KPS control room. The methodology used to evaluate the control room and offsite doses resulting from the FHA is consistent with RG 1.183 in conjunction with TEDE radiological units and limits, ARCON96 based onsite atmospheric dispersion factors, PAVAND based EAB and LPZ atmospheric dispersion factors, and Federal Guidance Reports No. 11 and 12 dose conversion factors. Isolation of the control room prior to movement of recently irradiated fuel assemblies and manual operator action to initiate the CRPARS within 20 minutes of the release will be new requirements.

Amendment 190 for Kewaunee Power Station, dated March 8, 2007 (Reference 14),

represents the current licensing basis for the FHA.

3.3.1 FHA Scenario Description The design basis scenario for the radiological analysis of the FHA assumes that cladding damage has occurred to all of the fuel rods in one dropped fuel assembly. The rods are assumed to instantaneously release their fission gas contents to the water surrounding the fuel assembly. The analyses include the evaluation of FHA cases that occur in both the containment and the spent fuel pool (SFP). All radioactivity released from the damaged fuel is released over a two hour period. Release pathways considered include:

1. Spent fuel pool via the Aux. Bldg stack

2. Spent fuel pool via the roll-up doors

3. Containment personnel hatch to the Aux Bldg stack

4. Containment to the Reactor Building stack

5. Containment equipment hatch

Serial Number 11-025A Page 73 of 191 A single KPS FHA scenario models the bounding FHA which does not credit mitigating systems (e.g., radiation monitor isolation, bypass and closure signals, or ventilation filtration) and maximizes source term, dispersion and dose. This bounding scenario provides the basis to allow all penetrations to be open under administrative control while moving recently irradiated fuel.

The results of this analysis show that control room isolation is required prior to moving recently irradiated fuel assemblies in order to maintain operator dose within 5 rem TEDE. KPS is proposing to remove credit for the Control Room Ventilation radiation monitor R-23 providing control room isolation. The R-23 system is not safety grade and consists of a single radiation monitor. In addition, the isolation signal generated by R-23 will not assure the closure of all control room ventilation dampers needed to provide complete control room isolation. The current Fuel Handling Accident (FHA) uses and credits the R-23 radiation monitor for control room isolation. The basis behind crediting R-23 relies on arguments that Operations will take appropriate actions to isolate the control room if R-23 fails to perform its isolation function. Removing credit for R-23 requires an alternative means to ensure control room isolation. The FHA requires that the control room be isolated prior to moving recently irradiated fuel and that manual operator action be taken to initiate the CRPARS within 20 minutes of the release.

3.3.2 FHA Source Term Definition In accordance with Regulatory Position 3 of RG 1.183, the core source term was previously calculated using the ORIGEN2 code for a Stretch Power Uprate (SPU) to 1772 megawatt thermal (MWt) and used in Amendment No. 172, issued February 27, 2004 (Reference 11). The core curies include a 6% increase to account for fuel management variations (493.6 +/- 10% EFPD, average enrichment of 4.5 w/o +/- 10%, and core mass of 49.1 MTU +/- 10%). This core inventory is described in the LOCA scenario (Section 3.2.2) and is used for the FHA with 100-hours of decay.

Serial Number 11-025A Page 74 of 191 For the FHA analyses, the core inventory was used to calculate the gap activity of one fuel assembly for input to RADTRAD-NAI. The amount of fuel damage is the same whether the FHA is in the spent fuel pool or containment. Therefore, the only variable between a FHA in the containment or spent fuel pool is the release point. As with previous AST submittals, the FHA analysis assumes the resulting chemical form of the radioiodine in the water is 99.85% elemental iodine and 0.15% organic iodide.

3.3.3 FHA Release Transport The FHA scenario does not credit operability or operation of the Spent Fuel Pool Sweep System nor does it credit any ventilation filtration systems or automatic functions. It is assumed that containment penetrations, (e.g., personnel hatch, equipment hatch, or other penetration) remain open for the duration of the 2-hour release. Modeling the release with the highest estimated control room X/Q from all possible release points and all possible intake points (normal intake and inleakage locations) maximizes the control room dose and represents the worst source to receptor orientation.

Releases from a FHA to the environment are at a rate of 3.454 air changes per hour.

This assures that greater than 99.9% of the activity is released within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . In addition, the release rate is conservatively biased to release > 80% of all activity within the first half hour of the event. No credit is taken for dilution or mixing of the activity released to the Auxiliary Building or Containment air volumes.

All possible release pathways were considered from a FHA in either the SFP or containment. The most conservative pathway to the Control Room was modeled. The bounding pathway is an unfiltered release from the Reactor Building Ventilation Exhaust Stack which has the largest calculated control room X/Q (see section 3.1.1). The EAB and LPZ dispersion factors encompass all possible release points (see Section 3.1.2),

and therefore are bounding.

Serial Number 11-025A Page 75 of 191 3.3.4 FHA Atmospheric Dispersion Factors 3.3.4.1 FHA Control Room X/Qs As described in Section 3.1, the onsite atmospheric dispersion factors were calculated using the ARCON96 code (Reference 5) and guidance from Regulatory Guide 1.194 (Reference 6). The FHA Control Room X/Qs listed in Table 1.3-4 were calculated for the following applicable KPS source points:

Reactor Building Exhaust Stack

Containment Equipment Hatch

Auxiliary Building Exhaust Stack

Fuel Area Roll-up Door 3.3.4.2 FHA Offsite (EAB & LPZ) /Qs As described in Section 3.1, the Exclusion Area Boundary (EAB) and Low Population Zone (LPZ) atmospheric dispersion factors were revised and are listed in Table 1.3-3.

These offsite atmospheric dispersion factors were generated using the PAVAND code (Reference 7) and guidance from Regulatory Guide 1.145 (Reference 8).

3.3.5 FHA Analysis Assumptions & Key Parameter Values 3.3.5.1 Method of Analysis The RADTRAD-NAI code (Reference 3) is used to calculate the radiological consequences from airborne releases resulting from a FHA at Kewaunee Power Station (KPS) to the EAB, LPZ, and Control Room.

RADTRAD can model a variety of processes that can attenuate and/or transport radionuclides. The RADTRAD models used in the FHA calculations include the following:

The damaged fuel assembly has been operating at the highest fuel rod power level. This conservative assumption maximizes fuel gap activity and dose consequences.

Serial Number 11-025A Page 76 of 191

All fuel rods in the dropped assembly fail, instantaneously releasing activity contained in the fuel gap into the water that the assembly is being moved within.

The overall pool decontamination factor (DF) for iodine is 200 (see Section 3.3.5.3).

25% of the fuel rods in the worst peak assembly do not comply with footnote 11 in RG 1.183 (Reference 1). Higher gap fractions applicable to the FHA and previously approved in KPS License Amendment 190 on March 8, 2007 (Reference 14) are assumed in these rods (see Table 3.3-1).

75% of the fuel rods in the worst peak assembly meet the criteria in RG 1.183 footnote 11 and use the suggested gap fractions for non-LOCA events.

Iodine leaving the water is 57% elemental and 43% organic per RG 1.183. All noble gases release instantaneously to the air above the water.

All activity released from the water surface is released to the environment within a 2-hour period without credit for mixing or dilution within the building volume.

The maximum X/Qs from any applicable release point to either the control room intake or control room inleakage pathway is used throughout the entire 2-hour release.

The FHA approved in Amendment 190 (Reference 14) differs from the FHA in this amendment request by the following:

1. Revised Control Room X/Qs (based on ARCON96)

2. Revised Off-site X/Qs (based on PAVAND)

3. Control Room Inleakage Assumption decreased from 1500 cfm to 800 cfm

4. Percent of fuel rods in the dropped assembly that exceed the criteria set forth in footnote 11 of RG 1.183 decreased from 50% to 25%

5. Automatic isolation of the control room by the non safety-grade, non-redundant control room ventilation monitor R-23 is no longer credited

6. Control room isolation is required prior to moving recently irradiated fuel assemblies

Serial Number 11-025A Page 77 of 191

7. All spent fuel pool area and containment penetrations (including the equipment hatch) are allowed to be open under administrative control during fuel manipulations

8. CRPARS is credited for operation within 20 minutes of the FHA based on operator action The combined effect of these changes result in changes to the EAB, LPZ, and control room doses due to a KPS design basis FHA. In all cases, the doses fall within required limits.

Figure 3.3-1 displays the assumptions, inputs and pathways used in RADTRAD to model the KPS FHA.

3.3.5.2 Basic Data and Assumptions Changes have been made to the AST FHA. Table 3.3-1 provides a complete list of inputs and assumptions used to reanalyze the KPS FHA.

Serial Number 11-025A Page 78 of 191 Figure 3.3-1 RADTRAD Model for FHA Containment or Spent Fuel Pool, 1 ft3 Environment Control Room 127,600 ft 3

800 cfm unfiltered in-leakage 800 cfm out-leakage 3.454 air changes/ hour 0-20 min: 0 cfm filtered recirculation

> 20 min: 2250 cfm filtered recirculation 90% elemental iodine filter efficiency 90% organic iodine filter efficiency 99% particulate iodine filter efficiency

Serial Number 11-025A Page 79 of 191 Table 3.3-1 Basic Data and Assumptions for FHA Parameter or Assumption CLB Value Proposed Value Reason for Change Source Term Fuel Damage:

1 assembly No Change Decay Time:

100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months No Change Radial Peaking Factor:

1.7 No Change Duration of Release 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months No Change Pool Decontamination Factor:

Noble Gases: 1 Iodines: 200 (effective DF)

No change Percentage of Fuel Rods that Exceed the Requirements of Footnote 11 of RG 1.183 50%

25%

Excess margin is being removed from analysis. This limit is reflected in the KPS blank Reload Safety Analysis Checklist and verified on a cycle specific basis. For rods above footnote 11 criteria, the gap fractions listed in Regulatory Guide 1.25 (as modified by the direction of NUREG/CR-5009) are used with the design peaking factor of 1.7.

Serial Number 11-025A Page 80 of 191 Table 3.3-1 Basic Data and Assumptions for FHA Parameter or Assumption CLB Value Proposed Value Reason for Change Gap Fractions Fuel that complies with footnote 11 of Regulatory Guide 1.183 Fuel that does not comply with footnote 11 of Regulatory Guide 1.1.83 I-131 0.08 Kr-85 0.10 Other 0.05

- noble gases

- halogens I-131 0.12 Kr-85 0.30 Other 0.10

- noble gases

- halogens No Change No Change Release Point:

Not applicable (One site control room X/Q represented any release point to the control room)

Reactor Building Exhaust Stack Current control room X/Q is treated as the bounding X/Q from any release point to the control room.

New ARCON96 analyses (Table 1.3-4) have been performed to analyze control room X/Qs from all release points.

Serial Number 11-025A Page 81 of 191 Table 3.3-1 Basic Data and Assumptions for FHA Parameter or Assumption CLB Value Proposed Value Reason for Change Activity in One Fuel Assembly Nuclide Activity (Ci)

I-131 2.99E+05 I-132 2.53E+05 I-133 3.15E+04 I-135 2.25E+01 Kr-85m 2.22E-02 Kr-85 4.72E+03 Kr-87 4.75E-19 Kr-88 7.77E-06 Xe-131m 4.50E+03 Xe-133m 1.06E+04 Xe-133 5.75E+05 Xe-135m 3.60E+00 Xe-135 1.10E+03 No Change EAB X/Q (sec/m3) 0 - 2 hr 2.232E-04 1.76E-04 New PAVAND X/Q values (see Table 1.3-3 and Section 3.1.2)

LPZ X/Q (sec/m3)

Period LPZ 0 - 2 hr 3.977E-05 2 - 24 hr 4.100E-06 1 - 2 day 2.427E-06 2 - 30 day 4.473E-07 Period LPZ 0 - 8 hr 3.36E-05 8 - 24 hr 2.37E-05 1 - 4 day 1.12E-05 4 - 30 day 3.94E-06 New PAVAND X/Q values (see Table 1.3-3 and Section 3.1.2)

Serial Number 11-025A Page 82 of 191 Table 3.3-1 Basic Data and Assumptions for FHA Parameter or Assumption CLB Value Proposed Value Reason for Change Control Room Control Room Volume (ft3) 127,600 No Change Normal Ventilation Unfiltered Makeup Air Flow (scfm) 2,750 No Change Filtered Recirculation Air Flow (scfm) 2,250 No Change Open Penetrations Personnel Hatch ANY penetration will be allowed to be open under administrative control during movement of irradiated fuel LAR request to allow ANY penetration to be open under Administrative Control CRPARS Filter Efficiency

(%)

Elemental Organic Particulate 90 (includes safety factor of 2) 90 (includes safety factor of 2) 99 No Change Control Room Unfiltered Inleakage (cfm) 1500 800 Maximum (ASTM) E741 tracer gas test = 447+/-51 cfm (Ref.

20)

Serial Number 11-025A Page 83 of 191 Table 3.3-1 Basic Data and Assumptions for FHA Parameter or Assumption CLB Value Proposed Value Reason for Change Control Room X/Q (sec/m3) 2.93E-3 4.88E-03 New ARCON96 estimates of control room X/Q (Table 1.3-4) show the Rx Bldg Exhaust Stack has the highest dispersion factor to the control room of any applicable release pathway Control Room Isolation (min) 1 0

CR ventilation intake rad monitor R-23 is no longer credited. Open penetration allowance will require the control room to be isolated prior to movement of recently irradiated fuel.

Control Room Post Accident Recirculation system (CRPARS) Start (min) 1 20 Operator action is required to start CRPARS within 20 minutes of event initiation based on communication with the refuel operator and radiation monitors going into alarm.

Serial Number 11-025A Page 84 of 191 3.3.5.3 Decontamination Factor in Less than 23 feet of Water Per Regulatory Guide 1.183, if the depth of water above a damaged fuel assembly is 23 feet or greater, the decontamination factors (DF) for elemental and organic species are 500 and 1, respectively, giving an overall effective DF of 200. Design configuration of a fuel assembly drop in the containment and spent fuel pool where examined to confirm the water depth of 23 feet. Based on the assumption that the fuel assembly will be horizontal once it comes to rest, it was determined that an assembly lying on the reactor vessel flange will have approximately 22.35 feet of water above the highest point of the assembly to the water surface. In the spent fuel pool, greater than 23 feet of water will exist.

The depth of 22.35 feet of water was evaluated to verify an effective decontamination factor of 200 using WCAP-7828 (Reference 27). Using the methods defined in the WCAP with conservative assumptions to minimize predicted decontamination factors for various depths of water, a DF of greater than 500 was determined for elemental iodine.

The use of an overall effective DF of 200 was determined to be appropriate per RG 1.183.

3.3.5.4 Recently Irradiated Fuel Determination The age of Recently Irradiated Fuel (RIF) was determined using an iterative approach to determine a decay time that results in a control room dose within the 5 rem limit without requirements for operability of control room emergency ventilation systems. Off-site dose analyses are unaffected by the determination of RIF. 375 hours0.00434 days 0.104 hours 6.200397e-4 weeks 1.426875e-4 months was selected as the basis for the definition of RIF based on RADTRAD runs that were made to determine when control room dose is < 5 rem TEDE without crediting any control room emergency ventilation or operator action. The worst case dispersion factor of any applicable release pathway (i.e., Reactor Building Exhaust Stack X/Q = 4.88E-03 sec/m3) was used in the control room dose model. The source term was determined by decaying the 100-hour decayed source term (net activity) from Table 3.3-1 by an

Serial Number 11-025A Page 85 of 191 additional 275 hours0.00318 days 0.0764 hours 4.546958e-4 weeks 1.046375e-4 months (for a total decay of 375 hours0.00434 days 0.104 hours 6.200397e-4 weeks 1.426875e-4 months ). The 375-hour decayed isotopic inventory used in the RADTRAD NIF file is listed below.

Nuclide Net Activity (Ci)

I-131 8.61E+01 I-133 1.99E-03 Kr-85 1.22E+03 Xe-133m 3.37E+01 Xe-133 1.52E+04 Recently Irradiated Fuel definition will be based on 375 hours0.00434 days 0.104 hours 6.200397e-4 weeks 1.426875e-4 months of decay, post-shutdown.

The control room operator dose based on RIF results in less than 5 rem TEDE.

3.3.6 FHA Analysis Results The offsite and control room doses are listed below. The KPS Fuel Handling Accident releases essentially all activity of one damaged fuel assembly over a two-hour release period.

The associated worst case TEDE for the FHA scenario is presented in Table 3.3-2. All doses are less than the limits specified in Regulatory Guide 1.183 and 10 CFR 50.67.

Table 3.3-2 Dose Summary for the Fuel Handling Accident Analysis Location TEDE (rem)

Limits (rem)

EAB 0.6 6.3 LPZ 0.2 6.3 Control Room 4.3 5

Serial Number 11-025A Page 86 of 191 3.4 Steam Generator Tube Rupture Accident This section describes the methods employed and the results of the Steam Generator Tube Rupture (SGTR) design basis radiological analysis. This analysis included doses associated with the releases of the radioactive material initially present in primary liquid, secondary liquid and iodine spiking. Doses are calculated at the Exclusion Area Boundary (EAB) for the worst-case two-hour period, at the Low Population Zone Boundary (LPZ), and in the KPS Control Room. The methodology used to evaluate the doses resulting from a SGTR is consistent with RG 1.183 (Reference 1) and utilized Federal Guidance Reports (FGR) No. 11 and 12 dose conversion factors.

3.4.1 SGTR Scenario Description A steam generator tube rupture (SGTR) is a break in a tube carrying primary coolant through the steam generator. This postulated break allows primary liquid to leak to the secondary side of one of the steam generators (denoted as the affected generator) with an assumed release to the environment through the steam generator Power Operated Relief Valves (PORVs). The PORV on the affected steam generator is assumed to open to control steam generator pressure at the beginning of the event, and remain open until operator action is taken to close the PORV within 55 minutes. Hence, the affected generator discharges steam to the environment for 55 minutes (0.92 hours0.00106 days 0.0256 hours 1.521164e-4 weeks 3.5006e-5 months ) until the generator is isolated by closure of the steam generator PORV. Flashed and un-flashed break flow in the affected steam generator is assumed to continue for the duration of the 55 minute period.

The intact generator discharges steam for a period of 29 hours3.356481e-4 days 0.00806 hours 4.794974e-5 weeks 1.10345e-5 months until the primary system has cooled sufficiently to allow a switchover to Residual Heat Removal System (RHRS) cooling. At 29 hours3.356481e-4 days 0.00806 hours 4.794974e-5 weeks 1.10345e-5 months , the RHRS can remove all the decay heat to achieve cold shutdown and steaming is no longer required for cooldown. No fuel damage is predicted as a result of a SGTR. Therefore, consistent with the current licensing analysis basis, the SGTR analysis was performed assuming both a pre-accident iodine spike and a concurrent accident iodine spike. In addition, the impact of a coincident

Serial Number 11-025A Page 87 of 191 loss-of-offsite power (LOOP) at the time of tube rupture was considered. In accordance with Regulatory Guide 1.183, release of noble gases without credit for holdup has been analyzed.

3.4.2 SGTR Source Term Definition Initial radionuclide concentrations in the primary and secondary systems for the SGTR accident are determined based on the maximum Technical Specification levels of activity. The SGTR accident analysis indicates that no fuel rod failures occur as a result of these transients. Thus, radioactive material releases were determined by the radionuclide concentrations initially present in primary liquid, secondary liquid, and iodine spiking. These values are the starting point for determining the curie input for the RADTRAD-NAI code runs.

Regulatory Guide 1.183 specifies that the released activities should be the maximum allowed by the Technical Specifications. Table 3.4-1 lists all the primary and secondary liquid radionuclide concentrations that are used in the analysis. Primary side concentrations are based on the proposed new Technical Specification 3.4.16 limits of 16.4 Ci/gm DE Xe-133 for gross gamma and 0.1 Ci/gm DE I-131 for iodine.

Secondary side concentrations are based on the proposed new Technical Specification 3.7.16 limit of 0.05 Ci/gm DE I-131 for iodine. In addition, since there is not a Technical Specification limit for the secondary side gross gamma activity, activities in the steam generator liquid were derived by assuming one half of the primary side activity to ensure that a suitably conservative source term was used.

Serial Number 11-025A Page 88 of 191 Table 3.4-1 Primary Coolant and Secondary Side Nuclide Concentrations Nuclide Primary (Ci/gm)

Secondary (Ci/gm)*

Kr-85m 4.76E-02 Kr-85 2.37E-01 Kr-87 3.12E-02 Kr-88 9.04E-02 Xe-131m 8.40E-02 Xe-133m 9.49E-02 Xe-133 6.67E+00 Xe-135m 1.38E-02 Xe-135 2.40E-01 Xe-138 1.73E-02 Br-83 2.51E-03 1.26E-03 Br-84 1.24E-03 6.20E-04 I-130 9.49E-04 4.75E-04 I-131 7.82E-02 3.91E-02 I-132 7.97E-02 3.99E-02 I-133 1.17E-01 5.85E-02 I-134 1.62E-02 8.10E-03 I-135 6.40E-02 3.20E-02 Rb-86 8.95E-04 4.48E-04 Rb-88 1.13E-01 5.65E-02 Rb-89 5.15E-03 2.58E-03 Cs-134 8.17E-02 4.09E-02 Cs-136 9.10E-02 4.55E-02 Cs-137 6.02E-02 3.01E-02 Ba-137m 5.70E-02 2.85E-02 Cs-138 2.65E-02 1.31E-02

Serial Number 11-025A Page 89 of 191 Table 3.4-1 Primary Coolant and Secondary Side Nuclide Concentrations Nuclide Primary (Ci/gm)

Secondary (Ci/gm)*

Cr-51 5.40E-03 2.70E-03 Mn-54 4.00E-04 2.00E-04 Fe-55 2.10E-03 1.05E-03 Fe-59 5.10E-04 2.55E-04 Co-58 1.40E-02 7.00E-03 Co-60 1.30E-03 6.50E-04 Sr-89 1.15E-04 5.75E-05 Sr-90 5.73E-06 2.87E-06 Sr-91 1.54E-04 7.70E-05 Sr-92 3.43E-05 1.72E-05 Y-90 1.60E-06 8.00E-07 Y-91m 8.31E-05 4.16E-05 Y-91 1.55E-05 7.75E-06 Y-92 2.97E-05 1.49E-05 Y-93 9.85E-06 4.93E-06 Zr-95 1.80E-05 9.00E-06 Nb-95 1.80E-05 9.00E-06 Mo-99 2.10E-02 1.05E-02 Tc-99m 1.95E-02 9.75E-03 Ru-103 1.54E-05 7.70E-06 Ru-106 5.22E-06 2.61E-06 Rh-103m 1.53E-05 7.65E-06 Rh-106 5.22E-06 2.61E-06 Te-125m 1.91E-05 9.55E-06 Te-127m 8.64E-05 4.32E-05 Te-127 3.64E-04 1.82E-04

Serial Number 11-025A Page 90 of 191 Table 3.4-1 Primary Coolant and Secondary Side Nuclide Concentrations Nuclide Primary (Ci/gm)

Secondary (Ci/gm)*

Te-129m 2.94E-04 1.47E-04 Te-129 3.82E-04 1.91E-04 Te-131m 6.91E-04 3.46E-04 Te-131 3.73E-04 1.87E-04 Te-132 8.10E-03 4.05E-03 Te-134 7.91E-04 3.96E-04 Ba-140 1.15E-04 5.75E-05 La-140 3.88E-05 1.94E-05 Ce-141 1.76E-05 8.80E-06 Ce-143 1.34E-05 6.70E-06 Ce-144 1.33E-05 6.65E-06 Pr-143 1.69E-05 8.45E-06 Pr-144 1.33E-05 6.65E-06

Secondary equals primary times 0.5.

Regulatory Guide 1.183 stipulates that SGTR accidents consider iodine spiking above the value allowed for normal operations based both on a pre-accident iodine spike and a concurrent accident spike. For KPS, the maximum iodine concentration that will be allowed by the proposed Technical Specification 3.4.16 as the result of an iodine spike is 10 Ci/gm DE I-131. The spike limit is being lowered commensurate with the reduction in reactor coolant activity. The pre-accident iodine spike concentrations corresponding to 10 Ci/gm DE I-131 are listed in Table 3.4-2. Regulatory Guide 1.183 defines a concurrent iodine spike as an accident initiated value 335 times the appearance rate corresponding to the Technical Specification 3.4.16 limit for normal operation (0.1 Ci/gm DE I-131 RCS TS limit) for a period of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . The concurrent iodine spike appearance rates based on 335 times the 0.1 Ci/gm DE I-131

Serial Number 11-025A Page 91 of 191 concentration are listed in Table 3.4-3. Appearance rates developed address the issues raised by NSAL-00-004 (Reference 28).

The dose conversion factors used to calculate the TEDE doses and DE I-131 for the Steam Generator Tube Rupture accident were taken from Table 3.2-3 for the isotopes required by Regulatory Guide 1.183 for the SGTR analysis.

Table 3.4-2 Pre-accident Iodine Spike RCS Concentration Nuclide Iodine Activity in RCS 0.1 DE I-131 Ci/gm Iodine Activity in RCS 10 DE I-131 Ci/gm I-131 7.82E-02 7.82E+00 I-132 7.97E-02 7.97E+00 I-133 1.17E-01 1.17E+01 I-134 1.62E-02 1.62E+00 I-135 6.40E-02 6.40E+00 Table 3.4-3 Concurrent Iodine Spike SGTR RCS Concentration Nuclide Appearance rate for 0.1 Ci/gm DE I-131 Ci/hr Spike = 335 SGTR Appearance Rate Ci/hr I-131 1.80E+00 6.02E+02 I-132 4.75E+00 1.59E+03 I-133 3.10E+00 1.04E+03 I-134 1.93E+00 6.45E+02 I-135 2.26E+00 7.57E+02

Serial Number 11-025A Page 92 of 191 3.4.3 SGTR Release Transport Affected Steam Generator The source term resulting from the radionuclides in the primary system coolant and from the iodine spiking in the primary system is transported to the affected steam generator by the break flow. The break flow is terminated after 55 minutes when the generator is isolated by closure of the PORV. A fraction of the break flow is assumed to flash to steam in the affected generator and to pass directly into the steam space of the affected generator with no credit taken for scrubbing by the steam generator liquid. The radionuclides entering the steam space as the result of flashing pass directly to the environment through the Steam Generator PORVs. The remainder of the break flow enters the steam generator liquid. Releases of radionuclides initially in the steam generator liquid and those entering the steam generator from the break flow are released as a result of secondary liquid boiling. A partition factor of 100 for all non-noble gas isotopes is assumed during boiling. Thus 1% of the iodines and particulates are released from the steam generator liquid to the environment along with the steam flow (moisture carryover is not actually modeled but is instead bounded by application of the partitioning factor). All noble gases are released from the primary system to the environment without reduction or mitigation. The transport model utilized for iodine and particulates was consistent with Appendix E of Regulatory Guide 1.183.

Intact Steam Generator The source term resulting from the radionuclides in the primary system coolant and from the iodine spiking in the primary system is transported to the intact generator by the leak-rate Limiting Condition for Operation (150 gallons per day) specified in Technical Specification LCO 3.4.13. All radionuclides in the primary coolant leaking into the intact generator are assumed to enter the steam generator liquid. Releases of radionuclides initially in the steam generator liquid and those entering the steam generator from the leakage flow are released as a result of secondary liquid boiling, including an allowance for a partition factor of 100 for all non-noble gas isotopes. Thus 1% of the iodines and particulates are assumed to pass into the steam space and then directly to the

Serial Number 11-025A Page 93 of 191 environment. All noble gases that are released from the primary system to the intact generator are released to the environment without reduction or mitigation. Releases were assumed to continue from the intact generator for a period of 29 hours3.356481e-4 days 0.00806 hours 4.794974e-5 weeks 1.10345e-5 months after which the RHRS is credited for removing 100% of decay heat with no requirement for steaming to augment cooldown.

3.4.4 SGTR Atmospheric Dispersion Factors 3.4.4.1 SGTR Control Room /Qs As described in Section 3.1, the onsite atmospheric dispersion factors were calculated using the ARCON96 code (Reference 5) and guidance from Regulatory Guide 1.194 (Reference 6). The SGTR Control Room X/Qs listed in Table 1.3-4 were calculated for the following applicable KPS source points:

A Steam Generator PORV

B Steam Generator PORV The control room X/Qs represent the highest values calculated based on the shortest distance measured from each applicable source location to control room receptor location (see Figure 3.1-1).

3.4.4.2 SGTR Offsite (EAB & LPZ) /Qs As described in Section 3.1, the Exclusion Area Boundary (EAB) and Low Population Zone (LPZ) atmospheric dispersion factors were revised and are listed in Table 1.3-3.

These offsite atmospheric dispersion factors were generated using the PAVAND code (Reference 7) and guidance from Regulatory Guide 1.145 (Reference 8).

3.4.5 SGTR Key Analysis Assumptions and Inputs 3.4.5.1 Method of Analysis The RADTRAD-NAI code (Reference 3) is used to calculate the radiological consequences from airborne releases resulting from a SGTR at Kewaunee Power Station (KPS) to the EAB, LPZ, and Control Room.

Serial Number 11-025A Page 94 of 191 There are several aspects of the SGTR analysis that require multiple RADTRAD models due to limitations of the code. This is due primarily to treatment of the source terms because noble gases are released without mitigation and iodines and particulates are released crediting partitioning and moisture carryover. The different models include:

Pre-incident spike impact - iodine and daughters

Pre-incident spike impact - noble gas

Coincident spike impact - iodine and daughters

Coincident spike impact - noble gas

Initial RCS TS activity - iodine and particulate

Initial RCS TS activity - noble gas

Secondary side bulk liquid - iodine and particulate A schematic shown in Figure 3.4-1 provides an overall picture of the SGTR releases to environment. Maximum and minimum values are provided for secondary side bulk liquid mass. The minimum value is used to reduce holdup for primary to secondary releases and the maximum value is used to maximize secondary side inventory; this is done to maximize dose from primary to secondary side releases.

3.4.5.2 Basic Data & Assumptions for SGTR The Basic Data and Assumptions are listed below in Table 3.4-4. A time-line of events is provided in Table 3.4.5. Steam and break flow data are listed in Tables 3.4-6 to 3.4-

8. Control room information is available in both Tables 3.4-4 and 1.3-1.

Serial Number 11-025A Page 95 of 191 Figure 3.4-1 SGTR Radioactive Release Schematic Liquid Break Flow Rate Hr lbm/ min 0 4680 0.0481 4332 0.92 0

Primary-to-Secondary leak rate (150 gpd)

Hr lbm/ min 0 0.869 29 0.0 Flashed Break Flow Rate Hr lbm/ min 0

1166 0.0481 756 0.92 0

RCS Mass = 262,735 lbm Unaffected SG liquid 84,000 lbm min 97,064 lbm max Unaffected SG Steam Affected SG liquid 84,000 lbm min 97,064 lbm max Affected SG Steam Iodine, particulates, and progeny released via partitioning and carryover to SG steam.

Hr lbm/ min 0

647 0.0481 32 0.92 0

Iodine, particulates, and progeny released via partitioning and carryover to SG steam.

Hr lbm/ min 0

647 0.0481 20 2

14 8

7 29 0

Affected SG Steam Release to the environment Hr lbm/ min 0

64668 0.0481 3186 0.92 0

RCS noble gases & iodine progeny (XE) released without mitigation Unaffected SG Steam Release to the environment Hr lbm/ min 0

64668 0.0481 1992 2

1356 8

690 29 0

RCS noble gases & iodine progeny (XE) released without mitigation (150 gpd)

Serial Number 11-025A Page 96 of 191 Table 3.4-4 Basic Data and Assumptions for SGTR Parameter or Assumption CLB Value Proposed Value Reason for Change Source Term Primary Coolant Specific Activity Limit DE I-131 (Ci/gm)

Gross Activity 1

Not Included 0.1

0.1 Ci/gm DE I-131 Technical Specification limits were reduced in order to maintain control room doses within acceptable limits.

Derived from the 1% failed fuel inventory and equivalent to the failed fuel for the TS DE I-131 limit.

Primary Coolant Concentrations at TS Limit Ci/gm I-131 I-132 I-133 I-134 I-135 7.80E-01 7.93E-01 1.16E+00 1.61E-01 6.37E-01 7.82E-02 7.97E-02 1.17E-01 1.62E-02 6.40E-02 Current values include 5% variation to consider minor variations in fuel design (e.g., enrichment, core mass and cycle length). Proposed values are adjusted to allow 10% variation, to make consistent with similar allowance built into core inventory curies.

Primary Coolant Noble Gas Activity 595 Ci/gm DE Xe-133 16.4 Ci/gm DE Xe-133 The revised Noble gas limit corresponds to an equivalent level of fuel failure (0.027%) as the TS DE I-131 limit of 0.1 Ci/gm Iodine Spike 500 335 Per RG 1.183 Accident-Initiated spike Duration (hr) 4 8

Per RG 1.183

Serial Number 11-025A Page 97 of 191 Table 3.4-4 Basic Data and Assumptions for SGTR Parameter or Assumption CLB Value Proposed Value Reason for Change Iodine Appearance Rate I-131 I-132 I-133 I-134 I-135 Ci/min 0.301 0.788 0.519 0.319 0.377 Ci/hr 1.80 4.75 3.10 1.93 2.26 The difference in the iodine appearance rates reflects a unit conversion and a factor of ten reduction directly related to the reduced TS specific activity limit.

Values on a consistent unit basis are shown to the right.

Ci/min 0.030 0.079 0.052 0.032 0.038 Primary to Secondary Leak rate (gpd/SG)*

150 No Change Pre-Accident Spike Coolant Activity (Ci/gm DE I-131) 20 10 Proposed TS spike limit was lowered commensurate with primary coolant activity reduction.

Iodine Partitioning PC for iodine = 100 No Change Iodine chemical form of Primary-to-Secondary Leakage (%)

Elemental 97 Organic 3

Particulate 0 No Change Moisture Carryover in Unaffected Steam Generators 1%

No Change Tube Uncovery.

No tube bundle uncovery assumed.

No Change Scrubbing of Flashed Break Flow Not Credited No Change

Serial Number 11-025A Page 98 of 191 Table 3.4-4 Basic Data and Assumptions for SGTR Parameter or Assumption CLB Value Proposed Value Reason for Change Secondary Iodine Activity Concentration 0.1 Ci/gm DE I-131 0.05 Ci/gm DE I-131 Proposed TS change SGTR Parameters Reactor Trip Time (sec) 173.3 No Change Safety Injection Signal (sec) 173.3 185 Conservative value based on the Westinghouse T&H analysis Operator Action to isolate Affected SG (min) 30 55 Conservative value confirmed in Operator timing studies Action to Align RHRS (hr) 24 29 Conservative assumption that RHRS start is delayed to 29 hrs.

Release to Environment (hr)

Unaffected SG Affected SG 0 - 24 0 - 0.5 0 - 29 0 - 0.92 Conservative assumption that RHRS start is delayed to 29 hrs.

Time for operator action to close PORV was increased to 55 minutes Reactor coolant mass (gm) 1.19E+08 No Change Initial Steam Generator Liquid Mass (lbm/SG) 84,000 - Min vol. used to minimize hold-up in the SG 97,064 - Max vol. used to maximize secondary side activity No Change No Change

Serial Number 11-025A Page 99 of 191 Table 3.4-4 Basic Data and Assumptions for SGTR Parameter or Assumption CLB Value Proposed Value Reason for Change Tube Rupture Break Flow (lbm)

Pre-Trip Post-Trip 16,900 {5,850 lbm/min}

138,000 {5,088 lbm/min}

No Change 265,200 The time assumed to close the PORV increased from 30 minutes to 55 minutes. Conservatively, the break flow rate at 30 minutes is assumed to persist for an additional 25 minutes.

Tube Rupture Break Flow Flashing Fraction Pre-Trip Post Trip 0.1993 0.1476 No Change Steam Release (lbm/min)

Ruptured SG Pre-Trip Post-Trip Intact SG Pre-Trip Trip - 2 hr 2 - 8 hr 8 - 24 hr 24 - 29 hr 6.47E+04 3.19E+03 6.47E+04 1.99E+03 1.36E+03 6.90E+02 0

No Change No Change No Change No Change No Change No Change 6.90E+02 RHR cut-in time was increased to 29 hours3.356481e-4 days 0.00806 hours 4.794974e-5 weeks 1.10345e-5 months . The steam release rate from 8 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months was conservatively maintained for an additional 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months .

Serial Number 11-025A Page 100 of 191 Table 3.4-4 Basic Data and Assumptions for SGTR Parameter or Assumption CLB Value Proposed Value Reason for Change Release points Ruptured SG Pre-Trip Post-Trip Intact SG Pre and Post-Trip Condenser SG Power Operated Relief Valves (PORVs)

PORVs PORVs PORVs The current analysis credits the condenser until reactor trip.

The revised analysis assumes Loss of Offsite Power (LOOP) coincident with the accident. All releases from the ruptured and intact SG will release through the PORVs.

Operator Action to close Affected SG PORV (min) 30 55 Conservative value confirmed in Operator timing studies EAB X/Q (sec/m3) 0 - 2 hr 2.232E-04 1.76E-04 New PAVAND X/Q values (see Table 1.3-3 and Section 3.1.2)

LPZ X/Q (sec/m3)

Period LPZ 0 - 2 hr 3.977E-05 2 - 24 hr 4.100E-06 1 - 2 day 2.427E-06 2 - 30 day 4.473E-07 Period LPZ 0 - 8 hr 3.36E-05 8 - 24 hr 2.37E-05 1 - 4 day 1.12E-05 4 - 30 day 3.94E-06 New PAVAND X/Q values (see Table 1.3-3 and Section 3.1.2)

Serial Number 11-025A Page 101 of 191 Table 3.4-4 Basic Data and Assumptions for SGTR Parameter or Assumption CLB Value Proposed Value Reason for Change Control Room Control Room Isolation (sec) 300 195 Current value was reduced to remove conservatism. Revised Control Room isolation includes SI signal at 185 seconds + 10 seconds for Control Room Damper closure.

Control Room Post Accident Recirculation System (CRPARS) Ventilation (sec) 300 318 CRPARS initiation includes SI signal at 185 seconds + 10 seconds for Diesel start + 63 seconds for Diesel sequencing + 60 seconds for CRPARS damper to open Control Room Unfiltered Inleakage (cfm) 1000 800 Maximum (ASTM) E741 tracer gas test = 447+/-51 cfm (Ref. 20)

Control Room HVAC Parameters (cfm)

Unfiltered Inleakage Unfiltered Make-up Air Filtered Recirculation 0 - 300 s 300 s - 30 d 0

1000 2750 0

0 2250 0 - 195 s 195 - 318 s 318 - 30 d 0

800 800 2750 0

0 0

0 2250 Unfiltered inleakage is not assumed until the control room is isolated at 195 seconds. Inleakage is assumed at 800 cfm, consistent with other DBA analyses.

Control Room Volume (ft3) 127,600 No Change Normal Ventilation Unfiltered Makeup Air (scfm) 2,750 No Change Filtered Recirculation Air Flow (scfm) 2,250 No Change

Serial Number 11-025A Page 102 of 191 Table 3.4-4 Basic Data and Assumptions for SGTR Parameter or Assumption CLB Value Proposed Value Reason for Change CRPARS Filter Efficiency

(%)

Elemental Organic Particulate 90 (includes safety factor of 2) 90 (includes safety factor of 2) 99 No Change Control Room X/Q (sec/m3) 0 - 8 h 2.93E-3 8 - 24 h 1.73E-3 1 - 4 d 6.74E-4 4 - 30 d 1.93E-4 Affected Intact B SG A SG 0 - 0.055 h 7.92E-3 2.24E-3 0.055 - 2 h 5.84E-3 2.46E-3 2 - 8 h 2.34E-2 2.13E-3 8 - 24 h 8.67E-3 8.60E-4 1 - 4 d 6.97E-3 6.96E-4 4 - 30 d 6.41E-3 5.81E-4 NEW ARCON96 X/Q values B SG 0 - 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months values from Table 1.3-4 have been reduced by a factor of 5 due to plume rise (see sec. 3.4.5.3). A SG PORVs have a horizontal exhaust, therefore plume rise reduction for the A SG X/Qs cannot be made.

Prior to CR isolation (0.055 h) the X/Q is to the CR intake. Post isolation, the X/Q represents the worst CR inleakage pathway into the turbine building.

The density used to convert volumetric leak rates (gpd) to mass leak rates (lbm/hr) was consistent with the basis of surveillance tests used to show compliance with leak rate technical specifications.

Serial Number 11-025A Page 103 of 191 3.4.5.3 SGTR Plume Rise Determination Following the guidance of RG 1.194, the buoyant plume rise associated with energetic releases from steam relief values or atmospheric steam dumps can be credited if (1) the release is uncapped and vertical, and (2) the time-dependent vertical velocity exceeds the 95th percentile wind speed, at the release point height, by a factor of 5.

The 95th percentile wind velocity was determined using meteorological data from 2002-2006. The value of the 95th percentile 10 meter and 60 meter wind speeds was found to be 7.6 and 11.6 meters per second, respectively. The B steam generator PORV has a larger atmospheric dispersion factor than the A PORV because of the close proximity of B PORV to the control room intake and turbine building intake locations. The steam flow from the B PORV is vertical and uncapped at the point where it enters the atmosphere. The elevation at which the steam enters the atmosphere is 6821 or 23.34 meters above grade. Using linear interpolation, the 95th percentile wind speed at this elevation is 8.6 meters per second. Five times this speed is 43 meters per second.

With a PORV exhaust stack cross sectional area of 2.02 square feet, the flow from an open PORV would need to equal or exceed 632 lbm/min* to equal an exit velocity of 43 meters per second. From Table 3.4-4, the steam flow from the affected steam generator exceeds 632 lbm/min for the entire accident duration. For conservatism, only the 0-2 hour X/Q for the B (Affected) SG PORV release is reduced by a factor of five, crediting the plume rise reduction allowed by RG 1.194.

(conservatively assumed at atmospheric pressure saturated steam conditions)

Serial Number 11-025A Page 104 of 191 Table 3.4-5 Time Line of Events Time, post accident Event seconds hours 0

0 SGTR - PORV sticks open LOOP 173.3 0.0481 Reactor Trip 185 0.0514 SI Actuated 195 0.0542 Control Room Isolates 318 0.0883 CRPARS initiated 3,300 0.92 PORV Closed (Affected SG Release Terminated) 104,400 29 RHRS Placed In Service (Intact SG Release Terminated) 2,592,000 720 Event Terminated Table 3.4-6 RCS Break Flow to Affected Steam Generator Time period (hour)

Total Break Flow Rate Flashed Break Flow Rate Liquid Break Flow Rate From To (lbm/min)

(lbm/min)

(lbm/min) 0 0.0481 5850 1166 4684 0.0481 0.92 5088 756 4332 0.92 720 0

0 0

Table 3.4-7 Affected Steam Generator Steam Release to Environment Time period (sec)

Time period (hour)

Release Rate From To From To (lbm/min) 0 173.3 0

0.0481 64,668 173.3 3300 0.0481 0.92 3,186 3300 2,592,000 0.92 720 0

Serial Number 11-025A Page 105 of 191 Table 3.4-8 Intact Steam Generator Steam Release to the Environment Time period, sec Time period, hour Release Rate From To From To (lbm/min) 0 173.3 0

0.0481 64,668 173.3 7,200 0.0481 2

1,992 7,200 28,800 2

8 1,356 28,800 86,400 8

24 690 86,400 104,400 24 29 690 104,400 2,592,000 29 720 0

3.4.6 SGTR Analysis Results The results of the analyses are presented in Table 3.4-9 for the Concurrent Spike and for the Pre-accident Iodine Spike.

Table 3.4-9 Dose Summary for the SGTR Accident Location TEDE (rem)

Limits (rem)

Concurrent Iodine Spike EAB 0.2 2.5 LPZ 0.1 2.5 Control Room 1.1 5

Pre-Accident Iodine Spike EAB 0.3 25 LPZ 0.1 25 Control Room 3.9 5

Serial Number 11-025A Page 106 of 191 3.5 Main Steam Line Break Analysis This section describes the methods employed and results of the Main Steam Line Break (MSLB) design basis radiological analysis. This analysis includes doses associated with the releases of radioactive material initially present in primary and secondary liquids at maximum allowable Technical Specification concentrations and adjusting for iodine spiking scenarios. No fuel failure is expected. Doses were calculated at the exclusion area boundary (EAB), at the low population zone (LPZ), and in the Control Room. The methodology used to evaluate the control room and offsite doses resulting from the MSLB accident is consistent with Regulatory Guide 1.183 (Reference 1) in conjunction with TEDE radiological units and limits, ARCON96 based onsite atmospheric dispersion factors, and Federal Guidance Report No. 11 and 12 (References 15 & 16, respectively) dose conversion factors.

3.5.1 MSLB Scenario Description The Main Steam Line Break (MSLB) accident begins with a break in one of the main steam lines leading from a steam generator (affected generator) to the turbine. Main steam line piping exits the containment and remains interior to the auxiliary building until entering into the turbine building. The control room is within this building matrix, with adjacent walls and entrances to both the auxiliary and turbine buildings. As discussed in Section 3.1.1, the primary pathway and assumed source of inleakage into the control room is through doors adjacent to the turbine building.

In order to determine maximum control room dose, both a steam line break in the turbine building and a break in the auxiliary building were separately evaluated (see Figures 3.5-1 and 3.5-2). This is a change from the current MSLB analysis which assumes the break releases directly into the atmosphere. Each evaluation considered conservative and bounding assumptions to determine which pathway scenario resulted in the maximum control room and offsite dose consequences.

Serial Number 11-025A Page 107 of 191 The worst case evaluated MSLB scenario for the EAB, LPZ and control room, involves a steam line break in the turbine building. This scenario will form the conditions, requirements and assumptions for the design basis MSLB accident.

MSLB in the Turbine Building The affected generator will dry out quickly and release all of the activity initially in the affected generator bulk liquid within 10 minutes directly into the turbine building where there are direct unfiltered inleakage pathways into the control room. Participation with 50% of the turbine building volume is credited for activity entering the building from the break. The pressure surge caused by the steam break will open turbine building blow-outs. During the assumed 10-minute initial release of steam generator (SG) contents, radioactive release from the blow-outs are set equal to the steam flow from the break.

Releases from the affected SG will continue after blow down due to primary-to-secondary leakage at the Technical Specification Limiting Condition for Operation 3.4.13.d rate of 150 gallons per day until the MSIV is closed by operator action, which was conservatively assumed to occur at 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . Radioactivity that escapes through blow-outs is modeled with both low and high volume release rates to negate any benefits. Loss of off-site power is assumed. As a result, the condenser is unavailable.

Cool down of the primary system is through the release of steam from the intact generator which is also assumed to have a primary-to-secondary leak at the Technical Specification rate of 150 gallons per day. Intact SG steaming will continue until sufficient cooldown at 29 hours3.356481e-4 days 0.00806 hours 4.794974e-5 weeks 1.10345e-5 months allows use of the residual heat removal system (RHRS).

In accordance with RG 1.183, Appendix E, two independent cases are evaluated. Case one assumes a pre-accident iodine spike, while the second case assumes a concurrent iodine spike.

3.5.2 MSLB Source Term Definition As with the SGTR accident, the analysis of the MSLB accident indicates that no fuel rod failures occur as a result of the transient. Thus, radioactive material releases are

Serial Number 11-025A Page 108 of 191 determined by assuming the radionuclide concentrations initially present in primary and secondary liquid at maximum Technical Specification limits and iodine spiking.

The Main Steam Line Break analysis uses the primary and secondary liquid source term discussed in Table 3.4-1 and the pre-accident iodine spike source term discussed in Table 3.4-2. The MSLB analysis also assumes a concurrent iodine spike listed below in Table 3.5-1 corresponding to an accident initiated value 500 times the appearance rate.

The appearance rate has decreased by a factor of ten from the current license basis values due to the proposed Technical Specification 3.4.16 RCS limit reduction for normal operation (0.1 Ci/gm DE I-131).

Table 3.5-1 Concurrent Iodine Spike MSLB RCS Concentration Nuclide Appearance rate for 0.1 Ci/gm DE I-131 Ci/hr Spike = 500 MSLB Appearance Rate Ci/hr I-131 1.80E+00 8.98E+02 I-132 4.75E+00 2.37E+03 I-133 3.10E+00 1.55E+03 I-134 1.93E+00 9.63E+02 I-135 2.26E+00 1.13E+03 3.5.3 MSLB Release Transport The source term resulting from activity in the primary system coolant and from iodine spiking in the primary system is transported to the SGs by the leak-rate limiting condition for operation of 150 gallons per day per SG specified in the Technical Specifications (TS LCO 3.4.13.d).

For the affected generator, the release pathway is assumed to be directly into the turbine building with no credit taken for holdup, partitioning or scrubbing by the SG liquid. Activity released from the break is assumed to participate with 50% of the turbine building volume. From the turbine building, the activity is assumed to leak into the

Serial Number 11-025A Page 109 of 191 control room as well as pass into the environment through pressure relief blow-outs located around the turbine building. A portion of the activity released through the blow-outs will disperse in the atmosphere and be pulled back into ventilation intakes and louvers and mix with activity residing in the turbine building. The affected generator will release activity into the turbine building until isolated (8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ) or until the primary side is cooled to 212°F (69.2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ). The operator action to isolate the affected SG is currently required to be completed within 10 minutes as part of isolating auxiliary feedwater, but is conservatively delayed for 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months to increase the dose consequences. In addition, no credit is taken for the steam line isolation signal that would close the affected MSIV based on SI coincident with HI-HI steam flow.

The affected SG transport model utilized for noble gases, iodine and particulates was consistent with Appendix E of Regulatory Guide 1.183. During the first 10 minutes post-trip, the affected SG is assumed to steam dry as a result of the MSLB, releasing all of the nuclides in the secondary coolant that were initially contained in the SG. During the first 8 or 69.2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , the primary coolant is also assumed to leak into the affected SG at the rate of 150 gpd with all activity released unmitigated. After 69.2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months the RCS will have cooled to below 212oF and the release via this pathway terminates. The primary-to-secondary leak rate path is terminated in 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months when operator action to isolate the affected SG is credited.

The intact SG is assumed to leak for 29 hours3.356481e-4 days 0.00806 hours 4.794974e-5 weeks 1.10345e-5 months until shutdown cooling is credited for decay heat removal. The primary-to-secondary technical specification leak rate limit of 150 gpd is assumed to maximize the release rate through the SG PORVs. The tube bundles of the intact SG remain covered during the release because of the availability of the Auxiliary Feedwater System. Releases of iodines and particulates are limited due to moisture carryover or partitioning. Releases of noble gases are assumed to occur directly to the environment without any mitigation or holdup.

There are several nuclide transport models associated with the intact SG. Together, they ensure proper accounting of iodine, particulates and noble gas releases. The

Serial Number 11-025A Page 110 of 191 intact model includes 2 RCS volumes each with the maximum technical specification source term (16.4 Ci/gm DE Xe-133 and 0.1 Ci/gm DE I-131), one for volume noble gas releases and one for iodine and particulate releases. The first volume has a pathway for releases of noble gas activity to the environment at 150 gpd, with 100 percent efficient iodine and particulate filtration. The transport to the environment of noble gases from the primary coolant and from iodine and particulate daughters released from the filters occurs without any mitigation or holdup.

The second RCS volume is used to model releases of radionuclides, which are initially in the intact SG liquid and those entering the SG from the primary to secondary leakage flow, as a result of secondary liquid boiling. Due to iodine partitioning and moisture carryover, 1% of the iodine and particulates in the SG bulk liquid are released to the environment at the steaming rate. The effect of partitioning and moisture carryover is modeled by reducing the steam flow rate by a factor of 100 to conserve radionuclides in the intact SG liquid. Radionuclides initially in the steam space do not provide any significant dose contribution and are not considered.

The pre-accident iodine spike is modeled in the same manner as the technical specification coolant activity model previously discussed.

The concurrent iodine spike model is modeled in the same manner as the technical specification coolant activity model but the iodine spike occurs for 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months after which the activity remaining in the primary coolant continues to be released for the remainder of the 29 hours3.356481e-4 days 0.00806 hours 4.794974e-5 weeks 1.10345e-5 months .

3.5.4 MSLB Atmospheric Dispersion Factors 3.5.4.1 MSLB Control Room /Qs As described in Section 3.1, the onsite atmospheric dispersion factors were calculated using the ARCON96 code (Reference 5) and guidance from Regulatory Guide 1.194

Serial Number 11-025A Page 111 of 191 (Reference 6). The MSLB Control Room X/Qs listed in Table 1.3-4 were calculated for the following applicable KPS release points:

A SG PORV

B SG PORV Control room X/Q values for the B SG PORV to the control room intake were selected to model release points applicable to the affected SG for a MSLB in the turbine building.

The X/Q is applied to releases from the turbine building blowouts. The control room intake will be isolated within 13 seconds of accident initiation (< 3 seconds for SI and 10 seconds for control room isolation damper closure). Therefore, any leakage into the control room will be from the turbine building (as discussed in Section 3.1.1) which has primary intake points near the South-West corner of the building. The B SG PORV to control room intake X/Q was selected because of its close proximity to the turbine building intake points and because this source-to-receptor combination results in the highest X/Q values. As shown in Figure 3.1-1, the B SG PORV is in close proximity to the turbine building Fan Room louvers. Table 1.3-4 shows that the B SG PORV X/Q values are the highest of any single release point at KPS. Uncorrected for plume rise, the B SG PORV X/Q values are conservative in comparison to an aggregate X/Q that would exist if calculated for multiple blow-outs located at various locations throughout the turbine building.

The A SG PORV X/Q values were used to model the intact SG releases. The 0-2 hour A SG PORV to control room intake X/Q value was used before isolation (13 seconds).

After isolation, the A SG PORV to turbine building fan room west louver X/Q values were used.

3.5.4.2 MSLB Offsite (EAB & LPZ) /Qs As described in Section 3.1, the Exclusion Area Boundary (EAB) and Low Population Zone (LPZ) atmospheric dispersion factors were revised and are listed in Table 1.3-3.

These offsite atmospheric dispersion factors were generated using the PAVAND code (Reference 7) and guidance from Regulatory Guide 1.145 (Reference 8).

Serial Number 11-025A Page 112 of 191 3.5.5 MSLB Key Analysis Assumptions and Inputs 3.5.5.1 Method of Analysis The RADTRAD-NAI code (Reference 3) is used to calculate the radiological consequences from airborne releases resulting from a MSLB at Kewaunee Power Station (KPS) to the EAB, LPZ, and Control Room.

RADTRAD can model a variety of processes that can attenuate and/or transport radionuclides during a MSLB. There are several aspects of the MSLB analysis that require multiple RADTRAD models due to limitations of the code. This is due primarily to treatment of the source terms and because noble gases are released without mitigation, and iodines and particulates are released crediting partitioning and moisture carryover in the intact SG, with no mitigation in the affected SG. The different models include:

Pre-incident spike - affected SG

Pre-incident spike - intact SG

Coincident spike - affected SG

Coincident spike - intact SG

Initial RCS TS activity - affected SG

Initial RCS TS activity - intact SG

Secondary side bulk liquid activity - affected SG

Secondary side bulk liquid activity - intact SG In order to determine maximum control room dose, both a steam line break in the turbine building and a break in the auxiliary building were evaluated (see Figures 3.5-1 and 3.5-2). The intact SG release is through the PORVs to the environment. Each evaluation considered conservative and bounding assumptions to determine which pathway scenario resulted in the maximum control room and offsite dose consequences. The worst case evaluated MSLB scenario for the EAB, LPZ and control

Serial Number 11-025A Page 113 of 191 room, involves a steam line break in the turbine building. This scenario will form the conditions, requirements and assumptions for the design basis MSLB accident.

The schematic shown in Figure 3.5-1 provides an overall picture of the design basis MSLB involving a break into the turbine building and releases to the environment.

Maximum and minimum values were used for secondary side bulk liquid mass. The minimum value is used to reduce holdup for primary to secondary releases in the intact SG and the maximum value is used to maximize secondary side inventory in the affected SG. This is done to maximize dose from primary to secondary side releases.

The evaluation of the break in the turbine building considered conservative and bounding assumptions to model the release from the affected SG into the turbine building volume. Participation with only 50% of the building volume was credited. Blow-out panels are assumed to open to relieve the pressure surge caused by the steam line break. High and low escape rates from the blow-outs (i.e., 10 building-volumes/hr down to 1 building-volume/hr) are modeled to maximize resulting control room and offsite dose consequences. Although multiple blow-out panels would open due to this event creating an aggregate X/Q to the control room, the highest control room X/Q values shown in Table 1.3-4 were used as conservative values to bound any release configuration that would result from a MSLB. The X/Q values used correspond to the B SG PORV release point that models a physical separation as close as 12 meters (see Table 3.1-1) between the PORV to the control room intake and turbine building louvers.

Evaluations for both the pre-accident and the concurrent iodine spike source terms were performed for a MSLB in the turbine building. Based on releases from the affected SG, which will persist for 69.2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months (the time necessary to cool the primary system down to 212°F), maximum dose consequences were determined for the EAB, LPZ and the control room. It became apparent that control room doses could not be maintained below 5 Rem for the concurrent iodine spike case if the affected generator releases primary coolant unmitigated into the turbine building at 150 gpd for the entire 69.2-hour

Serial Number 11-025A Page 114 of 191 cooldown period. Operator action is needed to isolate the affected SG within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months to maintain control room dose consequences within allowed limits. Existing Operation procedure steps (Reference 32) have the Operator closing the affected SG MSIV following a MSLB much earlier than 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months (< 10 minutes) in order to isolate feedwater flow. For evaluation purposes, an assumption to isolate the affected SG by closing the MSIV within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months was chosen to maximize consequences. This assumption also greatly relaxes the timing and burden on the operator to complete this action.

Steaming from the intact SG continues for 29 hours3.356481e-4 days 0.00806 hours 4.794974e-5 weeks 1.10345e-5 months until RCS pressure reduces to a level where RHRS can be used to remove decay heat.

In accordance with RG 1.183, Appendix E, two independent cases are evaluated. Case one assumes a pre-accident iodine spike, while the second case assumes a concurrent iodine spike. As previously discussed, the concurrent iodine spike case credits operator action to close the affected steam generator MSIV within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months accomplished through existing procedure actions (Reference 32). Conservatively, no action to close the MSIV was assumed for the pre-accident case to maximize dose consequences.

3.5.5.2 Basic Data & Assumptions for MSLB The Basic Data and Assumptions are listed in Table 3.5-2. Control room information is available in both Tables 3.5-2 and 1.3-1.

Serial Number 11-025A Page 115 of 191 Figure 3.5-1 MSLB Radioactive Release into the Turbine Building Schematic (Design Model)

Reactor Coolant System 262,736 lbm Control Room 127,600 ft3 CRPARS Recirculation 2250 cfm after automatic start Normal Control Room Intake 2750 cfm unfiltered until isolation Turbine Building

- 1.6E6 ft3 (50% of total)

Break flow of SG Direct Unfiltered inleakage 800 cfm Exhaust to Environment (equal to intake or inleakage)

Unfiltered inleakage from environment - 800 cfm -

after isolation Environment ISG Bulk Liquid & RCS Activity Release through PORVs or Safeties Environment ASG Bulk Liquid & RCS Activity Release ISG Primary-to-Secondary leak rate (150 gpd)

Hr lbm/min 0

0.869 29 0.0 {RHR}

ASG Primary-to-Secondary leak rate (150 gpd)

Hr lbm/min 0

0.869 69.2 0.0 {time to cool to 212 oF}

Affected Steam Generator (ASG) Liquid 161,000 lbm initial inventory plus feedwater released in 10 minutes Intact Steam Generator (ISG)

Liquid - no tube uncovery 84,000 lbm Credit for partitioning of iodine & moisture carryover of particulates to reduce the release ISG Noble Gas Primary-to-Secondary leak rate (150 gpd) to Environment

Serial Number 11-025A Page 116 of 191 Figure 3.5-2 MSLB Radioactive Release into the Auxiliary Building Schematic Unfiltered inleakage from environment - 800 cfm -

after isolation Reactor Coolant System 262,736 lbm Intact Steam Generator (ISG)

Liquid - no tube uncovery 84,000 lbm Credit for partitioning of iodine & moisture carryover of particulates to reduce the release ISG Primary-to-Secondary leak rate (150 gpd)

Hr lbm/min 0

0.869 29 0.0 {RHR}

Auxiliary Building

- East Quadrant 125,000 ft3 (50% of total)

ASG Bulk Liquid &

Primary Side Activity Release Control Room 127,600 ft3 CRPARS Recirculation 2250 cfm after automatic start Normal Control Room Intake 2750 cfm unfiltered until isolation Environment ISG Bulk Liquid & RCS Activity Release through PORVs or Safeties Pressurized Auxiliary Building Ducts 0-10 minutes 233 cfm (C11492)

Turbine Building 1.6E6 ft3 (50% of total)

Break flow of SG blowing down over 10 minutes followed by Turnover of 1 to 10 vol/hr, representing a range of natural circulation &

forced ventilation.

Direct Unfiltered inleakage 800 cfm Exhaust to Environment Exhaust to Environment (equal to intake or inleakage)

Environment RCS Activity Release through AB Exhaust Stack ASG Bulk Liquid to the Turbine building through doors on the 606 ft elev.

ASG Primary-to-Secondary leak rate (150 gpd)

Hr lbm/min 0

0.869 69.2 0.0 {time to cool to 212 oF}

Affected Steam Generator (ASG) Liquid 161,000 lbm initial inventory plus feedwater released in 10 minutes ISG Noble Gas Primary-to-Secondary leak rate (150 gpd) to Environment

Serial Number 11-025A Page 117 of 191 Table 3.5-2 Basic Data and Assumptions for MSLB Parameter or Assumption CLB Value Proposed Value Reason for Change Source Term Primary Coolant Specific Activity Limit DE I-131 (Ci/gm)

Gross Activity 1

Not included 0.1

0.1 Ci/gm DE I-131 Proposed Technical Specification limit change.

Derived from the 1% failed fuel inventory and equivalent to the failed fuel for the TS DE I-131 limit.

Primary Coolant Concentrations at TS Limit Ci/gm I-131 I-132 I-133 I-134 I-135 7.80E-01 7.93E-01 1.16E+00 1.61E-01 6.37E-01 7.82E-02 7.97E-02 1.17E-01 1.62E-02 6.40E-02 Current values include 5% variation to consider minor variations in fuel design (e.g., enrichment, core mass and cycle length). Proposed values are adjusted to allow 10% variation, to make consistent with similar allowance built into core inventory curies.

Primary Coolant Noble Gas Activity 1% fuel defects 16.4 Ci/gm DE Xe-133 The revised Noble gas limit corresponds to an equivalent level of fuel failure (0.027%) as the TS DE I-131 limit of 0.1 Ci/gm Accident Initiated (Concurrent) Iodine Spike 500 No Change Accident-Initiated (Concurrent) Spike Duration (hr) 4 8

Per RG 1.183

Serial Number 11-025A Page 118 of 191 Table 3.5-2 Basic Data and Assumptions for MSLB Parameter or Assumption CLB Value Proposed Value Reason for Change Iodine Appearance Rate I-131 I-132 I-133 I-134 I-135 Ci/min 0.301 0.788 0.519 0.319 0.377 Ci/hr 1.80 4.75 3.10 1.93 2.26 The difference in the iodine appearance rates reflects a unit conversion and a factor of ten reduction directly related to the reduced TS specific activity limit.

Values on a consistent unit basis are shown to the right.

Ci/min 0.030 0.079 0.052 0.032 0.038 Primary to Secondary Leak rate (gpd/SG)*

150 No Change Pre-Accident Spike Coolant Activity (Ci/gm DE I-131) 60 10 Proposed TS spike limit was lowered commensurate with primary coolant activity reduction.

Iodine Partitioning in Intact SG PC for iodine = 100 No Change Iodine chemical form of Primary-to-Secondary Leakage (%)

Elemental 97 Organic 3

Particulate 0 No Change Moisture Carryover in Intact SG 1%

No Change

Serial Number 11-025A Page 119 of 191 Table 3.5-2 Basic Data and Assumptions for MSLB Parameter or Assumption CLB Value Proposed Value Reason for Change SG Iodine Partition Factor Faulted SG Intact SG 1.0 0.01 No Change Secondary Iodine Activity Concentration 0.1 Ci/gm DE I-131 0.05 Ci/gm DE I-131 Proposed TS change MSLB Parameters Safety Injection Signal (sec) 0

<3 This time is based on high-high steam flow signal on the intact SG of 2.9 seconds. The SI signal actually comes in based on a low-low steam pressure on the affected SG <<3 seconds.

Operator Action to close Affected SG MSIV (hr)

NA 8

The concurrent iodine spike in the turbine building analysis requires closing the MSIV on the affected SG within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months to maintain resulting control room dose within GDC 19 limits.

Current analysis does not assume affect SG isolation Action to Align RHRS (hr) 24 29 Conservative assumption that RHRS start is delayed to 29 hrs.

Reactor coolant Mass (gm) 1.19E+08 (262,736 lbm)

No Change

Serial Number 11-025A Page 120 of 191 Table 3.5-2 Basic Data and Assumptions for MSLB Parameter or Assumption CLB Value Proposed Value Reason for Change Release to Environment (hr)

Unaffected SG Affected SG Pre-accident spike Concurrent spike 0 - 24 72 72 0 - 29 69.2 8

Conservative assumption that RHRS start is delayed to 29 hrs.

The current basis T&H analysis is cooldown to 212°F in 69.2 hr. Use of 69.2 hr is a reduction in conservatism.

Operator Action credited for the concurrent spike - utilizing existing procedure actions to isolate the affected SG.

Release of Initial Mass in Faulted Generator (min) 2 10 Validation of Operator actions shows isolation of feedwater to the affected SG will take up to 10 minutes.

Extending the release duration to 10 minutes ensures that all of the Curies are released from the affected SG and results in higher doses.

Faulted SG Steam Mass (lbm) 4759 No Change

Serial Number 11-025A Page 121 of 191 Table 3.5-2 Basic Data and Assumptions for MSLB Parameter or Assumption CLB Value Proposed Value Reason for Change Initial SG Liquid Mass (lbm)

Faulted SG Intact SG 156,254 84,000 - Min volume used to minimize hold-up in the SG No Change No Change Total Faulted SG mass = 161,000 lbm (Steam and Liquid mass)

Faulted SG Release (lbm) 0 - 2 min 2 - 10 min 10 - 30 min 1.61E+05 0

0 No Change 1.03E+05 (feedwater) 0 The initial mass from the faulted SG is released over 2 minutes.

Extending the period of release assures all activity that initially was in the SG is released.

Intact SG Release (lbm) 0 - 2 hr 2 - 8 hr 8 - 24 hr 24 - 29 hr 2.22E+05 4.24E+05 6.14E+05 0

No Change No Change No Change 1.92E+05 RHR cut-in time was increased to 29 hours3.356481e-4 days 0.00806 hours 4.794974e-5 weeks 1.10345e-5 months . The steam release rate from 8 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months was conservatively maintained for an additional 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months .

Turbine Building Volume (ft3) NA 3.19E+06 50% credit = 1.60E+06 ft3 Worst case MSLB occurs in the turbine building. Current basis assumes MSLB occurs into the environment.

Serial Number 11-025A Page 122 of 191 Table 3.5-2 Basic Data and Assumptions for MSLB Parameter or Assumption CLB Value Proposed Value Reason for Change Release points MSL Break (Affected SG)

Intact SG Environment Environment B SG is the affected generator releasing into the turbine building and released to the environment from blowout panels.

A SG PORV The current analysis assumes a break directly into the environment since the method utilized only one station control room X/Q value that was supposed to represent and bound all possible release points. A break into the turbine building is bounding over a break into the auxiliary building.

To maximize control room dose, the B SG is modeled as the affected SG. Therefore, the A SG PORV represents the intact SG.

Turbine Building Release to Environment (cfm)

NA 0 - 2 min 2.16E+06 2 - 10 min 3.45E+05 10 min - 30 d 2.67E+04 Affected SG blow down occurs into the turbine building (TB) in the first 2 minutes. From 2 to 10 minutes, until feedwater is isolated, steaming continues. After 10 minutes, primary to secondary releases from the affected SG continue into the TB.

One TB volume turnover per hour was assumed to maximize control room dose. Evaluation up to 10 volumes/hour (2.67E+05 cfm) showed decreasing control room doses and relative insensitivity of the offsite results to large changes in volumetric release rates.

Serial Number 11-025A Page 123 of 191 Table 3.5-2 Basic Data and Assumptions for MSLB Parameter or Assumption CLB Value Proposed Value Reason for Change EAB X/Q (sec/m3) 0 - 2 hr 2.232E-04 1.76E-04 New PAVAND X/Q values (see Table 1.3-3 and Section 3.1.2)

LPZ X/Q (sec/m3)

Period LPZ 0 - 2 hr 3.977E-05 2 - 24 hr 4.100E-06 1 - 2 day 2.427E-06 2 - 30 day 4.473E-07 Period LPZ 0 - 8 hr 3.36E-05 8 - 24 hr 2.37E-05 1 - 4 day 1.12E-05 4 - 30 day 3.94E-06 New PAVAND X/Q values (see Table 1.3-3 and Section 3.1.2)

Control Room Control Room Isolation (sec) 300 13 Current value was reduced to remove conservatism. Revised Control Room isolation includes SI signal at 3 seconds + 10 seconds for Control Room Damper closure.

Control Room Post Accident Recirculation System (CRPARS) Ventilation (sec) 300 136 CRPARS initiation includes SI signal at 3 seconds + 10 seconds for Diesel start + 63 seconds for Diesel sequencing + 60 seconds for CRPARS damper to open.

Control Room Unfiltered Inleakage (cfm) 1000 800 Maximum (ASTM) E741 tracer gas test = 447+/-51 cfm (Ref. 20).

Serial Number 11-025A Page 124 of 191 Table 3.5-2 Basic Data and Assumptions for MSLB Parameter or Assumption CLB Value Proposed Value Reason for Change Control Room HVAC Parameters (cfm)

Unfiltered Inleakage Unfiltered Make-up Air Filtered Recirculation 0 - 300 s 300 s - 30 d 0

1000 2750 0

0 2250 0 - 13s 13 - 136s 136s - 30 d 0

800 800 2750 0

0 0

0 2250 Unfiltered inleakage is not assumed until the control room is isolated at 13 seconds. Inleakage is assumed at 800 cfm, consistent with other DBA analyses.

Control Room Volume (ft3) 127,600 No Change Normal Ventilation Unfiltered Makeup Air Flow (scfm) 2,750 No Change Filtered Recirculation Air Flow (scfm) 2,250 No Change CRPARS Filter Efficiency

(%)

Elemental Organic Particulate 90 (includes safety factor of 2) 90 (includes safety factor of 2) 99 No Change Control Room X/Q (sec/m3) 0 - 8 h 2.93E-3 ASG ISG 0 - 2 h 3.96E-2 2.46E-3 2 - 8 h 3.20E-2 2.13E-3 8 - 24 h 1.21E-2 8.60E-4 NEW ARCON96 X/Q values Affected SG (ASG) utilizes X/Q values from Table 1.3-4 for the B SG PORV to the control room

Serial Number 11-025A Page 125 of 191 Table 3.5-2 Basic Data and Assumptions for MSLB Parameter or Assumption CLB Value Proposed Value Reason for Change 8 - 24 h 1.73E-3 1 - 4 d 6.74E-4 4 - 30 d 1.93E-4 1 - 4 d 1.01E-4 6.96E-4 4 - 30 d 8.58E-3 5.81E-4 intake to maximize CR dose. With only a 12 meter separation from the B PORV to the intake, the high X/Q bounds any aggregate X/Q that would result from modeling TB blowout panels to the nearest intake point.

Intact unaffected SG (ISG) releases from A PORV to the TB west louvers maximize the CR X/Q from the A SG. For the first 13 seconds, prior to CR isolation, the X/Q to the normal control room intake is 2.24E-3.

The density used to convert volumetric leak rates (gpd) to mass leak rates (lbm/hr) was consistent with the basis of surveillance tests used to show compliance with leak rate technical specifications.

Serial Number 11-025A Page 126 of 191 3.5.6 MSLB Analysis Results The total TEDE to the EAB, LPZ and Control Room from a Main Steam Line Break is summarized below in Table 3.5-3 for the concurrent and pre-accident spike. The concurrent spike results in the highest dose consequences for both offsite and the control room. All doses are within the limits specified in Regulatory Guide 1.183 and 10 CFR 50.67.

Table 3.5-3 Dose Summary for the MSLB Accident Location TEDE (rem)

Limits (rem)

Concurrent Iodine Spike EAB 0.1 2.5 LPZ 0.1 2.5 Control Room 4.2 5

Pre-Accident Iodine Spike EAB 0.1 25 LPZ 0.1 25 Control Room 4.7 5

Serial Number 11-025A Page 127 of 191 3.6 Locked Rotor Accident (LRA) Analysis This section describes the methods employed and results of the Locked Rotor Accident (LRA) design basis radiological analysis. The analysis assumes failure of 25% of the fuel rods, due to Departure from Nucleate Boiling (DNB) during the accident. Doses were calculated at the Exclusion Area Boundary (EAB), at the Low Population Zone (LPZ), and in the KPS control room. The methods used to evaluate the control room and offsite doses resulting from the LRA included Regulatory Guide 1.183 methodology, ARCON96-based control room atmospheric dispersion factors, PAVAND-based EAB and LPZ atmospheric dispersion factors, Federal Guidance Reports (FGR) No. 11 and 12 dose conversion factors, and credit for a new operator action to actuate the control room emergency ventilation system within one hour of the accident.

3.6.1 LRA Scenario Description The Locked Rotor Accident (LRA) begins with instantaneous seizure of a rotor in one of the two reactor coolant pumps. The sudden decrease in core coolant flow while the reactor is at power results in a degradation of core heat transfer that results in assumed fuel damage due to Departure from Nucleate Boiling (DNB). Although there is no increase in the leak rate of primary coolant to the secondary side during the LRA, a large amount of activity (from the failed fuel) is transported to the secondary side via any pre-existing leaks in the steam generators.

A turbine trip and coincident loss of offsite power are incorporated into the analysis.

This results in a release to the environment via power operated relief valves (PORV) with releases to the environment continuing until cooldown can be performed 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months post-accident using the Residual Heat Removal System (RHR). Operator action is credited for control room isolation and emergency ventilation actuation one hour following event initiation.

Kewaunee station is removing credit for the Control Room Ventilation Intake radiation monitor R-23 to provide control room isolation. The R-23 system is not safety grade

Serial Number 11-025A Page 128 of 191 and consists of a single radiation monitor. In addition, the isolation signal generated by R-23 is only a partial signal that will not assure the closure of all control room inlet and outlet ventilation dampers to provide complete control room isolation. Full isolation requires actions by the operator to close dampers that are not included in the isolation logic. The current Locked Rotor Accident (LRA) uses and credits the R-23 system for control room isolation. The basis behind the use of R-23 relies on arguments that Operations will take appropriate actions within 45 minutes to isolate the control room if R-23 fails to perform its isolation function. Removing credit for R-23 requires an alternative means to ensure control room isolation. Operator action will be required within one hour following a LRA to isolate the control room. One hour is sufficient time for the operator to identify the accident, take necessary emergency steps in response to the accident, and direct action to isolate the control room and start the control room emergency ventilation system. This new time-critical operator action will be incorporated into Operation procedures and validated.

3.6.2 LRA Source Term Definition The core source term used in the Locked Rotor Analysis is taken from Table 3.2-3.

Analyses are based on 25% of the gap activity being released, with gap activity based on Regulatory Position 3 of RG 1.183.

3.6.3 LRA Release Transport The release scenario uses the Technical Specification LCO 3.4.13.d primary to secondary leakage limit of 150 gpd per steam generator. The release from both steam generators continues for 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months until shutdown cooling can be placed into service to remove decay heat. After 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , the release from the steam generators is terminated.

The RADTRAD-NAI computer code (Reference 3) is used to model the time dependent transport of radionuclides, from the primary to secondary side and out to the environment via steam relief valves.

Serial Number 11-025A Page 129 of 191 3.6.4 LRA Atmospheric Dispersion Factors 3.6.4.1 LRA Control Room /Qs As described in Section 3.1, the onsite atmospheric dispersion factors were calculated using the ARCON96 code (Reference 5) and guidance from Regulatory Guide 1.194 (Reference 6). The LRA Control Room X/Qs listed in Table 1.3-4 were calculated for the following applicable KPS source points:

A Steam Generator PORV

B Steam Generator PORV The control room X/Qs determined represent the highest values calculated based on the shortest distance measured from each applicable source location to control room receptor location (see Figure 3.1-1).

3.6.4.2 LRA Offsite (EAB & LPZ) /Qs As described in Section 3.1, the Exclusion Area Boundary (EAB) and Low Population Zone (LPZ) atmospheric dispersion factors were revised and are listed in Table 1.3-3.

These offsite atmospheric dispersion factors were generated using the PAVAND code (Reference 7) and guidance from Regulatory Guide 1.145 (Reference 8).

3.6.5 LRA Analysis Assumptions and Key Parameters 3.6.5.1 Method of Analysis The RADTRAD-NAI code (Reference 3) is used to calculate the radiological consequences from airborne releases resulting from a LRA at Kewaunee Power Station (KPS) to the EAB, LPZ, and Control Room.

RADTRAD can model a variety of processes that can attenuate and/or transport radionuclides during a LRA. There are aspects of the LRA analysis that require two RADTRAD models due to limitations of the code. This is due primarily to treatment of the source terms because noble gases are released without mitigation and iodines and

Serial Number 11-025A Page 130 of 191 particulates are released crediting partitioning and moisture carryover. For conservatism, the postulated releases from assumed primary-to-secondary leakage of 150 gpd in each steam generator are combined and released from the generator PORV showing the highest control room X/Q value. The worst case release path for pre and post control room isolation is the B Steam Generator PORV. Note that the X/Q values for this pathway were reduced by a factor of 5 for the 0-2 hr and 2-8 hr periods. This reduction is taken following the guidance of RG 1.194, crediting the effects of plume rise for high velocity exhaust steam that exceeds the 95th percentile wind speed by a factor of 5 and adjusted for the physical release elevation. For explanation of this determination, see section 3.6.5.3.

A schematic shown in Figure 3.6-1 provides a summary of the LRA releases to environment.

Serial Number 11-025A Page 131 of 191 Figure 3.6-1 LRA Radioactive Release Schematic 3.6.5.2 Basic Data & Assumptions for LRA Changes have been made to the AST LRA. Table 3.6-1 provides a complete list of inputs and assumptions used to reanalyze the KPS LRA.

Primary-to-Secondary leak rate (150 gpd) x 2 SG Hr lbm/ min 0 1.74 29 0.0 RCS Mass = 262,735 lbm 25% Fuel Failure SG liquid 84,000 lbm/SG SG Steam Iodine, particulates, and progeny released via 0.01 partitioning to SG steam.

Hr lbm/ min 0

17.5 2

12.64 8

0 SG Steam Release to the environment Hr lbm 0-2 210,000 2-8 455,000 RCS noble gases & iodine progeny (XE) released without mitigation

Serial Number 11-025A Page 132 of 191 Table 3.6-1 Basic Data and Assumptions for LRA Parameter or Assumption CLB Value Proposed Value Reason for Change Source Term Primary to Secondary Leak rate (gpd/SG)*

150 No Change Failed Fuel Following the Accident (%)

50 25 Rods-in-DNB analysis show approximately 7% rods-in-DNB following a LRA for the current cycle. 25% is specified in the reload safety analysis checklist (RSAC).

Fraction of Core Activity in Gap (%)

I-131 Kr-85 Other Noble Gases Other Halogens Alkali Metals 8

10 5

5 12 No Change Iodine Partitioning PC = 100 No Change Alkali Metal Partitioning PC = 100 No Change Iodine chemical form of Primary-to-Secondary Leakage (%)

Elemental 97 Organic 3

Particulate 0 No Change Initial Secondary Side Coolant Activity Included Not Included Because fuel failure occurs, modeling of initial coolant activity is

Serial Number 11-025A Page 133 of 191 Table 3.6-1 Basic Data and Assumptions for LRA Parameter or Assumption CLB Value Proposed Value Reason for Change not required. CLB analysis shows less than 1% contribution to dose from secondary side activity.

Core Activity Table 3.2-3 No Change Radial Peaking Factor 1.7 No Change Tube Uncovery.

No tube bundle uncovery assumed.

No Change LRA Parameters RHR Cut-In Time (hr) 8 No Change Reactor Trip Time (sec) 0 No Change Loss of Offsite Power (sec) 0 No Change Safety Injection Signal None No Change Reactor coolant mass (gm) 1.19E+08 No Change Steam Generator Liquid Mass (lbm/SG) 0 - 30 minutes 30 min - 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 87,000 116,900 84,000 No Change Minimum SG liquid volume used to minimize hold-up

Serial Number 11-025A Page 134 of 191 Table 3.6-1 Basic Data and Assumptions for LRA Parameter or Assumption CLB Value Proposed Value Reason for Change Steam Release (lbm) 0 - 2 hr 2 - 8 hr 210,000 455,000 No Change Release point Not applicable (One site control room X/Q represented any release point to the control room)

B Steam Generator PORV New ARCON96 estimates of control room X/Q (Table 1.3-4) show the B PORV has the highest dispersion factor to the control room of any applicable release pathway.

EAB X/Q (sec/m3) 0 - 2 hr 2.232E-04 1.76E-04 New PAVAND X/Q values (see Table 1.3-3 and Section 3.1.2)

LPZ X/Q (sec/m3)

Period LPZ 0 - 2 hr 3.977E-05 2 - 24 hr 4.100E-06 1 - 2 day 2.427E-06 2 - 30 day 4.473E-07 Period LPZ 0 - 8 hr 3.36E-05 8 - 24 hr 2.37E-05 1 - 4 day 1.12E-05 4 - 30 day 3.94E-06 New PAVAND X/Q values (see Table 1.3-3 and Section 3.1.2)

Serial Number 11-025A Page 135 of 191 Table 3.6-1 Basic Data and Assumptions for LRA Parameter or Assumption CLB Value Proposed Value Reason for Change Control Room Control Room Volume (ft3) 127,600 No Change Control Room Isolation (min) 10.67 60 CLB credits control room intake radiation monitor R-23 to detect and isolate the control room. R-23 is not redundant and is no longer credited for CR isolation.

NEW Operator action is proposed to isolate the control room within 60 minutes of LRA utilizing multiple inputs as indicators of the accident (e.g., Rx coolant low flow and radiation monitor alarms).

Normal Ventilation Unfiltered Makeup Air Flow (scfm) 2,750 No Change Filtered Recirculation Air Flow (scfm) 2,250 No Change Control Room Post Accident Recirculation system (CRPARS) Ventilation (min) 11 60 CRPARS initiation is assumed to occur at 60 minutes, coincident with the operator action to isolate the control room.

Control Room Unfiltered Inleakage (cfm) 1500 800 Maximum (ASTM) E741 tracer gas test = 447+/-51 cfm (Ref. 20)

Serial Number 11-025A Page 136 of 191 Table 3.6-1 Basic Data and Assumptions for LRA Parameter or Assumption CLB Value Proposed Value Reason for Change Control Room HVAC Parameters (cfm)

Unfiltered Inleakage Unfiltered Make-up Air Filtered Recirculation 0 - 11 m 11 m - 30 d 0

1500 2750 0

0 2250 0 - 60 m 60 m - 30 d 0

800 2750 0

0 2250 Unfiltered inleakage is not assumed until the control room is isolated at 60 minutes at which time inleakage is assumed at 800 cfm, consistent with other DBA analyses.

CRPARS Filter Efficiency

(%)

Elemental Organic Particulate 90 (includes safety factor of 2) 90 (includes safety factor of 2) 99 No Change Control Room X/Q (sec/m3) 0 - 8 h 2.93E-3 8 - 24 h 1.73E-3 1 - 4 d 6.74E-4 4 - 30 d 1.93E-4 0 - 2 h 7.92E-3 2 - 8 h 6.40E-3 1.21E-2 1.01E-2 8.58E-3 0 - 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months values from Table 1.3-4 have been reduced by a factor of 5 due to plume rise (see Section 3.6.5.3).

Prior to CR isolation the X/Q is to the CR intake. Post isolation, the X/Q represents the worst CR inleakage pathway into the turbine building.

The density used to convert volumetric leak rates (gpd) to mass leak rates (lbm/hr) was consistent with the basis of surveillance tests used to show compliance with leak rate technical specifications.

Serial Number 11-025A Page 137 of 191 3.6.5.3 Plume Rise Determination Following the guidance of RG 1.194, the buoyant plume rise associated with energetic releases from steam relief valves or atmospheric steam dumps can be credited if (1) the release is uncapped and vertical, and (2) the time-dependent vertical velocity exceeds the 95th percentile wind speed, at the release point height, by a factor of 5.

The 95th percentile wind velocity was determined using meteorological data from 2002-2006. The value of the 95th percentile 10 meter and 60 meter wind speeds was found to be 7.6 and 11.6 meters per second, respectively. The B steam generator PORV has a larger atmospheric dispersion factor than the A PORV because of the close proximity to the control room intake and turbine building intake locations. The steam flow from the B PORV is vertical and uncapped at the point where it enters the atmosphere. The elevation at which the steam enters the atmosphere is 6821 or 23.34 meters above grade. Using linear interpolation the 95th percentile wind speed at this elevation is 8.6 meters per second. Five times this speed is 43 meters per second.

With a PORV exhaust stack cross sectional area of 2.02 square feet, the flow from an open PORV would need to equal or exceed 632 lbm/min* to equal an exit velocity of 43 meters per second. The steam flows for the LRA are 210,000 lbm for the first 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and 455,000 lbm for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . Because the steam release from the LRA is assumed from both generators, the single generator flow rates would be one half or 875 lbm/min for the first 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and 632 lbm/min from 2 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months following the LRA. For both the 0-2 hour and 2-8 hour periods following the LRA, flow out the PORV is sufficient to achieve an exhaust exit velocity that is five times higher than the 95th percentile wind speed. Therefore, the calculated ARCON96 control room X/Q values from the B PORV were reduced by a factor of 5 for the 0-2 and 2-8 hour periods.

(conservatively assumed at atmospheric pressure saturated steam conditions)

Serial Number 11-025A Page 138 of 191 3.6.6 LRA Results The results of the design basis Locked Rotor analysis are presented in Table 3.6-2.

These results show the calculated dose for the worst 2-hour interval (EAB), and for the assumed 30-day duration of the event for the control room and the LPZ. The doses are calculated with the TEDE methodology, and are compared with the applicable acceptance criteria specified in 10 CFR 50.67 and Regulatory Guide 1.183.

Table 3.6-2 TEDE Results for the Locked Rotor Accident Location TEDE (rem)

Limits (rem)

EAB 0.3 2.5 LPZ 0.2 2.5 Control Room 4.7 5

Serial Number 11-025A Page 139 of 191 3.7 RCCA Ejection Accident (REA) Analysis This section describes the evaluation of TEDE at the EAB, LPZ and Control Room from a KPS Rod Control Cluster Assembly (RCCA) Ejection Accident (REA). Two release cases are considered. The first case is a release into the containment. The second case is a release into the primary coolant, which is subsequently released through the secondary system.

3.7.1 REA Scenario Description This accident is defined as the mechanical failure of a control rod mechanism pressure housing, resulting in the ejection of a RCCA and drive shaft. The consequence of this mechanical failure is a rapid positive reactivity insertion in conjunction with an adverse core power distribution, possibly leading to localized fuel rod damage.

3.7.2 REA Source Term Definition The core source term used in the RCCA Ejection Accident Analysis are taken from Table 3.2-3. The release of the core source term is adjusted for the fraction of fuel rods assumed to fail during the accident and the fractions of core inventory assumed to be in the pellet-to-clad gap.

Less than 15 percent of the fuel rods in the core undergo DNB as a result of the rod-ejection accident. In determining the offsite doses following a rod-ejection accident, it is conservatively assumed that 15 percent of the fuel rods in the core suffer sufficient damage such that all of their gap activity is released. Ten percent of the total core activity of iodine and noble gases, and 12 percent of the total core activity for alkali metals are assumed to be in the fuel-cladding gap. In the calculation of activity releases from the failed/melted fuel, the maximum radial peaking factor of 1.7 was applied.

A small fraction of the fuel in the failed fuel rods is assumed to melt as a result of the rod ejection accident. This amounts to 0.375 percent of the core, and the melting takes place in the centerline of the affected rods. The 0.375 percent of the fuel assumes that

Serial Number 11-025A Page 140 of 191 15 percent of the rods in the core enter DNB. Of the rods that enter DNB, 50 percent are assumed to experience some melting of the fuel (7.5 percent of the core). Of the rods experiencing melting, 50 percent of the axial length of the rod is assumed to experience melting (3.75 percent of the core). It is further assumed that only 10 percent of the radial portion of the rod experiences melting (0.375 percent of the total core).

For both the containment leakage release path and the primary-to-secondary leakage release path, all noble gas and alkali metal activity released from the failed fuel (both gap activity and melted fuel activity) is available for release.

For the containment leakage release path, all of the iodine released from the gap of failed fuel and 25 percent of the activity released from melted fuel is available for release from containment.

The release fractions for both the containment and secondary system release scenarios were calculated as follows, using the design input and assumptions provided in Table 3.7-2.

Serial Number 11-025A Page 141 of 191 Input Description Value A

Radial Peaking Factor 1.7 B

Rods > DNB 15.0%

C

% Rods > DNB with centerline melt 50.0%

D

% inner rod melt limit 10.0%

E

% axial length with melt 50.0%

F Cesium Gap Fraction 12.0%

G Iodine and Noble Gas Gap Fractions 10.0%

H Fuel Noble Gas Available for Release 100.0%

I Fuel Iodine Available for Release to Containment 25.0%

J Fuel Iodine Available for Release to Secondary System 50.0%

K Cesium in Fuel Available for Release 100.0%

Percent of core fuel volume that is melted:

L = B

C

D

E = 0.375%

Percent of iodine core inventory released in the containment release scenario:

((G

B) + (I

L))

A = 2.71%

Percent of iodine core inventory released in the secondary side release scenario:

((G

B) + (J

L))

A = 2.87%

Noble gas release fraction used for both scenarios:

((G

B) + (H

L))

A = 3.19%

Cesium release fraction used for both scenarios:

((F

B) + (K

L))

A = 3.70%

Serial Number 11-025A Page 142 of 191 3.7.3 REA Release Transport Two release paths are considered for the REA: containment leakage and the secondary system.

The containment release transport assumptions and methodology are similar to the LOCA and can be found in section 3.2.5, with a few exceptions. The exceptions are:

1) The core release fractions are based on Appendix H of R.G. 1.183. The core release fractions are based on the consequences of 15% failed fuel and 0.375%

melted fuel.

2) Containment sprays do not initiate due to a REA. Therefore there are no consequences from ECCS leakage and RWST back-leakage.

3) The safety injection signal is initiated 4 minutes after a REA. Therefore, the control room is not isolated until 4 minutes 10 seconds following a REA.

The second release path is via the secondary system. The activity in the secondary system release is based on Appendix H of RG 1.183. The iodines released from the steam generators are assumed to be 97% elemental and 3% organic. The maximum allowable primary-to-secondary leak rate of 150 gpd per steam generator, which is specified in Technical Specification LCO 3.4.13.d, exists until shutdown cooling is in operation and release from the steam generators terminate. All noble gas radionuclides released to the secondary system are released to the environment without reduction or mitigation. The condenser is not available due to an assumed loss of offsite power. A partition coefficient for iodine of 100 is assumed in the steam generators.

The primary-to-secondary leak occurs during the first 30 minutes of the REA (until primary system pressure is less than secondary side system pressure). Steam generator mass releases are unchanged from previous Westinghouse thermal-hydraulic analyses. The steam released during the REA and subsequent cool-down is listed in Table 3.7-2.

Serial Number 11-025A Page 143 of 191 3.7.4 REA Atmospheric Dispersion Factors 3.7.4.1 REA Control Room /Qs As described in Section 3.1, the onsite atmospheric dispersion factors were calculated using the ARCON96 code (Reference 5) and guidance from Regulatory Guide 1.194 (Reference 6). The REA Control Room X/Qs listed in Table 1.3-4 were calculated for the following applicable KPS source points:

Reactor Building Exhaust Stack

Shield Building

Auxiliary Building Exhaust Stack

A Steam Generator PORV

B Steam Generator PORV The control room X/Qs determined represent the highest values calculated based on the shortest distance measured from each applicable source location to control room receptor location (see Figure 3.1-1).

3.7.4.2 REA Offsite (EAB & LPZ) /Qs As described in Section 3.1, the Exclusion Area Boundary (EAB) and Low Population Zone (LPZ) atmospheric dispersion factors were revised and are listed in Table 1.3-3.

These offsite atmospheric dispersion factors were generated using the PAVAND code (Reference 7) and guidance from Regulatory Guide 1.145 (Reference 8).

3.7.5 REA Analysis Assumptions and Key Parameters 3.7.5.1 Method of Analysis The RADTRAD-NAI code (Reference 3) is used to calculate the radiological consequences from the containment airborne release and primary-to-secondary release resulting from a REA at Kewaunee Power Station (KPS) to the EAB, LPZ, and Control Room.

Serial Number 11-025A Page 144 of 191 Table 3.7-1 shows the timing of events used in the REA analysis. The sequencing of events is derived from the design inputs and assumptions listed in Table 3.7-2.

Table 3.7-1 REA Event Timing Event Timing Statepoint Description Relationship (min)

(hr)

T0 Start of Event (instantaneous release) 0 sec 0

0 T1 SI Signal T0 + 240 sec 4

0.06667 T2 Control Room Isolation T1 + 10 sec 4.1667 0.06944 T3 CRPARS Starts T1 + 133 sec 6.2167 0.10361 T4 Shield Bldg Ventilation Starts T1 + 10 min 14 0.23333 T5 Secondary Side Releases Terminate T0 + 0.5 hr 30 0.5 T6 Shield Bldg Recirculation Starts T1 + 0.5 hr 34 0.56667 T7 Containment Leak Rates Decrease by 50%

T0 + 24 hr 1440 24 T8 End of Event T0 + 720 hr 43200 720 A schematic shown in Figure 3.7-1 provides a flowchart demonstrating the compartments and pathways used in RADTRAD to calculate the doses resulting from containment releases. Figure 3.7-2 provides a similar flowchart for secondary system releases resulting from a REA.

Serial Number 11-025A Page 145 of 191 Containment Leakage Model The primary containment leakage model assumes the failed fuel enters the containment and is released to the atmosphere through containment leakage. The natural deposition mechanism within the containment volume is modeled using the Powers model built into RADTRAD. Containment spray removal is not credited. The containment leaks directly to the environment, through the shield building, and through the auxiliary building special ventilation zone. Releases from the auxiliary building special ventilation zone to the environment are filtered. The shield building ventilation system filters the shield building air volume. A portion of the shield building air volume is discharged to the environment as necessary to maintain the negative pressure in the shield building annulus. Releases from the shield building to the environment are filtered. The shield building ventilation system fans establish a negative building pressure within the first 10 minutes after the safety injection signal. During that interval no credit is taken for filtering the shield building exhaust.

During the first 10 minutes of the accident, it is assumed that 90 percent of the activity leaking from the containment is discharged directly to the environment and 10 percent enters the Auxiliary Building where it is released through filters. After 10 minutes, only 1.0 percent of the activity leaking from the containment is assumed to go directly to the environment, 10 percent continues to go to the Auxiliary Building, and 89 percent is assumed to pass into the Shield Building. The air discharged from the Shield Building is filtered to remove iodine. Additionally, once the Shield Building is brought to subatmospheric pressure at 30 minutes into the event, the iodine is subject to removal by recirculation through filters. A shield building participation fraction of 0.5 is assumed.

Serial Number 11-025A Page 146 of 191 Figure 3.7-1 RADTRAD Model for Containment Airborne Releases

Serial Number 11-025A Page 147 of 191 Secondary System Model A secondary system release model assumes that 100% of the activity released from the fuel is completely dissolved in the primary coolant. This activity enters the secondary system via primary-to-secondary leakage and is then released to the environment.

During the first 250 seconds of the accident the control room is not isolated.

Figure 3.7-2 RADTRAD Model for Secondary System Releases 3.7.5.2 Basic Data & Assumptions for REA Changes have been made to the AST LRA. Table 3.7-2 provides a complete list of inputs and assumptions used to reanalyze the KPS LRA.

Serial Number 11-025A Page 148 of 191 Table 3.7-2 Basic Data and Assumptions for REA Parameter or Assumption CLB Value Proposed Value Reason for Change Source Term Core Power (MWt) 1782.6 (Licensed power of 1772 MWt with 0.6%

uncertainty)

No Change Core Inventory (curies)

Licensed Uprated Core based on 1782.6 MWt multiplied by 1.06 (Table 3.2-3)

No Change Gap Fraction (%)

Iodine Noble Gases Alkali Metals 10 10 12 No Change Initial Iodine Species in Containment (%)

Elemental Methyl (organic)

Particulate (aerosol) 4.85 0.15 95 No Change Rods in DNB (% of core) 15 No Change Melted Fuel (% of core) 0.375 No Change Power Peaking Factor 1.70 No Change

Serial Number 11-025A Page 149 of 191 Table 3.7-2 Basic Data and Assumptions for REA Parameter or Assumption CLB Value Proposed Value Reason for Change Activity Released into Containment (%)

From Failed Fuel From Melted Fuel Iodine Noble Gases Alkali Metals 100 25 100 100 No Change No change Activity Released into Primary Coolant (%)

From Failed Fuel From Melted Fuel Iodine Noble Gases Alkali Metals 100 50 100 100 No Change No change Initial Iodine Species in Containment (%)

Elemental Methyl (organic)

Particulate (aerosol) 4.85 0.15 95 No Change Primary-to-Secondary Leakage (gpd / SG)*

150 No Change

Serial Number 11-025A Page 150 of 191 Table 3.7-2 Basic Data and Assumptions for REA Parameter or Assumption CLB Value Proposed Value Reason for Change Iodine Species of Primary-to-Secondary Leakage

(%)

Elemental Methyl (organic)

Particulate (aerosol) 97 3

0 No Change Coolant Activity Concentration Prior to Accident Primary Iodine Primary Noble Gases Primary Alkali Metals Secondary Iodine Secondary Alkali 60 Ci/gm DE I-131 Equiv. to 1% fuel defects Equiv. to 1% fuel defects 0.1 Ci/gm DE I-131 10% of Primary conc.

NONE Per RG 1.183, appendix H, the source term for a PWR Rod Ejection accident only needs to consider the fuel damage postulated from the event.

The CLB analysis shows inclusion of activity prior to the accident contributes less than 1% to the overall consequences.

Reactor Coolant Mass (gm) 1.22E+08 1.19E+08 Because no initial activity is assumed, lower RCS mass is conservative (higher concentration of failed fuel activity).

Serial Number 11-025A Page 151 of 191 Table 3.7-2 Basic Data and Assumptions for REA Parameter or Assumption CLB Value Proposed Value Reason for Change Containment Containment Leak Rate (wt%/day) 0-24 hours

>24 hours 0.2 0.1 No Change Containment Leak Path Fractions 0-10 minutes Through Shield Bldg Through Aux Bldg SV Direct to Environment 10 minutes - 30 days Through Shield Bldg Through Aux Bldg SV Direct to Environment 0.0 0.10 0.90 0.89 0.10 0.01 No Change No Change Shield Building Drawdown Time: (Tech Specs) 10 minutes No Change Containment Volume (ft3) 1.32E6 No Change

Serial Number 11-025A Page 152 of 191 Table 3.7-2 Basic Data and Assumptions for REA Parameter or Assumption CLB Value Proposed Value Reason for Change Containment Spray and Iodine Removal Containment Spray Removal Not Credited No Change Natural deposition (hr-1)

Not Credited Powers Model set at the 10th percentile Per RG 1.183 Appendix H, natural deposition may be credited Shield Building Shield Building Annulus Volume (ft3) 3.74E+05 No Change Shield Building Participation Fraction 0.5 No Change Shield Building Ventilation and Recirculation Iodine Filter Efficiency (%)

Elemental Methyl (organic)

Particulate (aerosol) 90 90 99 95 (includes safety factor of 2) 95 (includes safety factor of 2) 99 Conservative filter efficiencies for elemental and organic iodine were increased to be consistent with other accident analyses. Safety factor of 2 remains.

Shield Building Air Flow to Environment (cfm) 0-10 min 10-30 min

>30 min 0

6600 3100 No Change

Serial Number 11-025A Page 153 of 191 Table 3.7-2 Basic Data and Assumptions for REA Parameter or Assumption CLB Value Proposed Value Reason for Change Shield Building Recirculation Flow (cfm) 0-30 min

>30 min 0

2300 No Change Auxiliary Building Participation with Auxiliary Building Volume or Hold-up None No Change Auxiliary Building Special Ventilation Iodine Filter Efficiency (%)

Elemental Methyl (organic)

Particulate (aerosol) 90 90 99 95 (includes safety factor of 2) 95 (includes safety factor of 2) 99 Conservative filter efficiencies for elemental and organic iodine were increased to be consistent with other accident analyses. Safety factor of 2 remains.

Secondary Release Primary to Secondary Leak rate (gpd from 2 SG)*

300 No Change Iodine Partitioning PC = 100 No Change Alkali Metal Partitioning PC = 100 No Change

Serial Number 11-025A Page 154 of 191 Table 3.7-2 Basic Data and Assumptions for REA Parameter or Assumption CLB Value Proposed Value Reason for Change Iodine chemical form of Primary-to-Secondary Leakage (%)

Elemental 97 Organic 3

Particulate 0 No Change Tube Uncovery.

No tube bundle uncovery assumed.

No Change Primary-to-Secondary Leak Duration (min) 30 No Change REA Parameters Safety Injection Signal (sec) 52.5 240 Delay of the SI signal is conservative. CLB assumption is based on a 2-inch diameter break.

The REA is specified to have a smaller 1.6 inch diameter break. SI signal generated from a 1-inch diameter break is 240 seconds.

Steam Generator Liquid Mass (lbm/SG) 87,000 84,000 Minimum SG liquid volume used to minimize hold-up consistent with other secondary system release accidents.

Steam Release to Environment (lbm/sec) 0 - 200 sec 200 - 1800 sec

>1800 sec 800 100 0

No Change

Serial Number 11-025A Page 155 of 191 Table 3.7-2 Basic Data and Assumptions for REA Parameter or Assumption CLB Value Proposed Value Reason for Change Release point(s)

Containment Pathway Secondary Release Containment / Shield Bldg Rx Building Stack Exhaust Aux Building Stack Exhaust B SG PORV No Change Release Termination (hr) 0.5 No Change EAB X/Q (sec/m3) 0 - 2 hr 2.232E-04 1.76E-04 New PAVAND X/Q values (see Table 1.3-3 and Section 3.1.2)

LPZ X/Q (sec/m3)

Period LPZ 0 - 2 hr 3.977E-05 2 - 24 hr 4.100E-06 1 - 2 day 2.427E-06 2 - 30 day 4.473E-07 Period LPZ 0 - 8 hr 3.36E-05 8 - 24 hr 2.37E-05 1 - 4 day 1.12E-05 4 - 30 day 3.94E-06 New PAVAND X/Q values (see Table 1.3-3 and Section 3.1.2)

Control Room Control Room Volume (ft3) 127,600 No Change

Serial Number 11-025A Page 156 of 191 Table 3.7-2 Basic Data and Assumptions for REA Parameter or Assumption CLB Value Proposed Value Reason for Change Normal Ventilation Unfiltered Makeup Air Flow (scfm) 2,750 No Change Filtered Recirculation Air Flow (scfm) 2,250 No Change Control Room Isolation (sec) 150 250 Control room isolation will occur 10 seconds following SI signal at 240 seconds.

Control Room Unfiltered Inleakage (cfm) 1000 800 To maintain consistency with all other radiological analyses.

Maximum (ASTM) E741 tracer gas test = 447+/-51 cfm (Ref. 20)

Control Room Post Accident Recirculation system (CRPARS) Ventilation (min) 150 373 CRPARS initiation is assumed to be effective 133 seconds following SI.

Based on 10 second delay to switchover from normal ventilation to emergency operation, 63 second delay in diesel loading of CRPARS, and 60 seconds to open recirculation dampers.

Serial Number 11-025A Page 157 of 191 Table 3.7-2 Basic Data and Assumptions for REA Parameter or Assumption CLB Value Proposed Value Reason for Change Control Room HVAC Parameters (cfm)

Unfiltered Inleakage Unfiltered Make-up Air Filtered Recirculation 0 - 150 s 150 s - 30 d 0

1000 2750 0

0 2250 0 - 250 s 250 s - 30 d 0

800 2750 0

0 2250 Unfiltered inleakage is not assumed until the control room is isolated at 250 seconds at which time inleakage is assumed at 800 cfm, consistent with other DBA analyses.

CRPARS Filter Efficiency

(%)

Elemental Organic Particulate 90 (includes safety factor of 2) 90 (includes safety factor of 2) 99 No Change Control Room X/Q (sec/m3)

Containment / Shield Bldg Rx Bldg Stack Exhaust Aux Bldg Stack Exhaust B SG PORV for all releases 0 - 8 hrs 2.93E-03 8 - 24 hrs 1.73E-03 24 - 96 hrs 6.74E-04 96 - 720 hrs 1.93E-04 CR Intake Inleakage 0 - 2 hr 0 - 2 hr 1.84E-03 1.74E-03 4.88E-03 3.97E-03 3.67E-03 2.90E-03 3.96E-02 2.92E-02 New ARCON96 control room X/Q estimates (Table 1.3-4)

Prior to CR isolation (250 sec) the X/Q is to the CR intake. Post isolation, the inleakage X/Q represents the worst CR inleakage X/Q, by way of the turbine bldg, from each respective release point.

For period values out to 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months , see Table 1.3-4

The density used to convert volumetric leak rates (gpd) to mass leak rates (lbm/hr) was consistent with the basis of surveillance tests used to show compliance with leak rate technical specifications.

Serial Number 11-025A Page 158 of 191 3.7.6 REA Analysis Results The total TEDE to the EAB, LPZ, and the Control Room from a RCCA Ejection Accident (REA) is summarized below in Table 3.7-3 for the containment and the secondary side release pathways. The containment pathway results in the highest dose consequences for both offsite and the control room. All doses are within the limits specified in Regulatory Guide 1.183 and 10 CFR 50.67.

Table 3.7-3 TEDE Results for the RCCA Ejection Accident Location TEDE (rem)

Limits (rem)

Containment Release Pathway EAB 0.2 6.3 LPZ 0.1 6.3 Control Room 0.8 5

Secondary Side Release Pathway EAB 0.1 6.3 LPZ 0.1 6.3 Control Room 0.5 5

Serial Number 11-025A Page 159 of 191 3.8 Waste Gas Decay Tank Analysis This section describes the methods employed and results of the Waste Gas Decay Tank failure (WGDT) design basis radiological analysis. The analysis assumes activity stored in a gas decay tank consists of the noble gases released from the processed coolant with only negligible quantities of the less volatile isotopes. Doses were calculated at the Exclusion Area Boundary (EAB), at the Low Population Zone (LPZ),

and in the KPS Control Room. The methodology used to evaluate the control room and offsite doses resulting from a WGDT accident included Standard Review Plan Branch Technical Position 11-5 (Reference 19), Regulatory Guide 1.24 (Reference 30),

ARCON96-based control room atmospheric dispersion factors, PAVAND-based EAB and LPZ atmospheric dispersion factors, and Federal Guidance Reports (FGR) No. 11 and 12 dose conversion factors.

The current WGDT analysis credits the control room post accident recirculation system (CRPARS) in the determination of control room dose. New analyses have been performed to demonstrate that the CRPARS ventilation system is not required to maintain control room dose within acceptable limits.

3.8.1 WGDT Scenario Description The Waste Gas Decay Tank (WGDT) accident is defined as an unexpected and uncontrolled release to the atmosphere of the radioactive xenon and krypton fission gases that are stored in the waste gas storage system. Failure of a gas decay tank or associated piping could result in a release of this gaseous activity. The activity in a gas decay tank is taken to be the maximum amount that could accumulate from operation with cladding defects in 1 percent of the fuel elements. Per the guidance in Regulatory Guide 1.24, all gaseous radioactive material is assumed to release to the atmosphere over a 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months period.

Serial Number 11-025A Page 160 of 191 3.8.2 WGDT Source Term Definition The activity assumed in the WGDT analysis, previously calculated for the stretch power uprate and approved in Amendment No. 172, (Reference 11), represents the maximum activity of noble gases, xenon and krypton, accumulated over a full core cycle with 1%

failed fuel. The maximum WGDT inventory for each nuclide is given in Table 3.8-1.

These activities are extremely conservative compared to actual activity which would accumulate in the gas decay tanks based on revised reactor coolant activity limits. New Technical Specification requirements will limit reactor coolant activity (see Sections 2.3 and 3.4.2) to a small fraction of current requirements. The source term represented in Table 3.8-1 does not include the associated reductions that would be caused by operating at reduced RCS limits.

Table 3.8-1 Waste Gas Decay Tank Activity (Ci)

Nuclide Activity in GDT Ci Kr-85m 8.53E+01 Kr-85 2.39E+03 Kr-87 1.58E+01 Kr-88 1.08E+02 Xe-131m 5.20E+02 Xe-133m 4.76E+02 Xe-133 3.85E+04 Xe-135m 2.78E+01 Xe-135 6.68E+02 Xe-138 1.84E+00 3.8.3 WGDT Release Transport The release scenario assumes the failure of a gas decay tank into the Auxiliary Building.

No credit is taken for building volume dilution. The radioactive content of the tank is assumed to release over a two hour period. The release is modeled using the Auxiliary

Serial Number 11-025A Page 161 of 191 Building Stack Exhaust as the release point to maximize the control room dose. The effluent resulting from the postulated event is assumed to release to the environment without continuous effluent radiation monitoring to automatically isolate and/or terminate the effluent release. No credit is taken for control room isolation, so the release is assumed to transport directly to the control room intake.

3.8.4 WGDT Atmospheric Dispersion Factors 3.8.4.1 WGDT Control Room /Qs As described in Section 3.1, the onsite atmospheric dispersion factors were calculated using the ARCON96 code (Reference 5) and guidance from Regulatory Guide 1.194 (Reference 6). The WGDT Control Room X/Qs listed in Table 1.3-4 were calculated for the following applicable KPS source point:

Auxiliary Building Stack Exhaust The control room X/Q determined for the Auxiliary Building Stack Exhaust to the Control Room Intake represents the highest value applicable to any source to receptor combination for the WGDT accident.

3.8.4.2 WGDT Offsite (EAB & LPZ) /Qs As described in Section 3.1, the Exclusion Area Boundary (EAB) and Low Population Zone (LPZ) atmospheric dispersion factors were revised and are listed in Table 1.3-3.

These offsite atmospheric dispersion factors were generated using the PAVAND code (Reference 7) and guidance from Regulatory Guide 1.145 (Reference 8).

3.8.5 WGDT Analysis Assumptions and Key Parameters 3.8.5.1 Method of Analysis The RADTRAD-NAI code (Reference 3) is used to calculate the radiological consequences from airborne releases resulting from a WGDT at Kewaunee Power Station (KPS) to the EAB, LPZ, and Control Room.

Serial Number 11-025A Page 162 of 191 The total contents of the WGDT are assumed to be released over a 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months period.

Assuming a removal rate of 3.45/hr will release essentially all (99.9%) of the gas decay tank contents over a two hour period. This equates to a release rate, for use in RADTRAD, of 8289%/day.

There are aspects of the WGDT analysis that require multiple RADTRAD models.

Conservative combinations of control room ventilation rates (filtered recirculation and unfiltered inleakage) and control room isolation times (varied from quick isolation to delayed isolation to no isolation) were modeled in order to maximize control room dose and prove the CRPARS is not needed to maintain control room dose within acceptable limits (see Section 3.8.5.3).

A schematic shown in Figure 3.8-1 provides picture summary of the WGDT release to the environment modeled in RADTRAD.

3.8.5.2 Basic Data & Assumptions for WGDT Changes have been made to the WGDT analysis. Table 3.8-2 provides a complete list of inputs and assumptions used to reanalyze the KPS WGDT event.

Serial Number 11-025A Page 163 of 191 Figure 3.8-1 WGDT Radioactive Release Schematic Auxiliary Building GDT volume set to 1 ft3 GDT release of 8289%/day 3.8.5.3 Assumptions to Maximize Control Room Dose Control room dose is calculated with a combination of control room assumptions that maximize control room dose (i.e., 30 minute delayed isolation of the control room in conjunction with low unfiltered inleakage). This combination of assumptions maximizes control room occupant exposure during a short (2-hour) duration release and bounds the condition of no control room isolation. Intake of the radioactive material at the maximum control room ventilation intake rate will achieve a delayed equilibrium concentration as flow is both into and out of the control room. Delayed isolation of the control room and reducing the intake/outflow combination will trap the radioactivity within the control room and maximize the exposure to the occupants. Time sensitivity runs determined 30 minutes as the time that would maximize the control room dose. An unfiltered inleakage rate of 200 cfm was assumed to maximize control room dose. This rate is lower than one half of the minimum unfiltered inleakage air flow of 409 +/- 29 cfm measured by the American Society for Testing and Materials (ASTM) E741 (tracer gas) leakage test conducted in December 2004 (Reference 20). Unfiltered inleakage rates greater than 200 cfm produce lower control room dose due to the associated purge effect of the inflow.

WGDT Table 3.8-1 1% FAILED FUEL Environment

Serial Number 11-025A Page 164 of 191 Table 3.8-2 Basic Data and Assumptions for WGDT Parameter or Assumption CLB Value Proposed Value Reason for Change Source Term WGDT Radiation Source (Curies)

Table 3.8-1 No Change Dose Consequence Multiplier 1.1 1.12 Source term adjustment factor to allow for fuel management variations. Previous multiplier allowed variation in cycle length of 493.6 +/- 5% EFPD. The new higher multiplier accounts for a larger fuel management variation, similar to that required in the RSAC of 493.6 +/-

10% EFPD.

RCS Coolant Activity

(% failed fuel) 1 No Change Core Activity Table 3.2-3 No Change WGDT Parameters Release Duration (min) 5 120 RG 1.24 allows the release period for the GDT rupture to be 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

Release Rate (%/day) 1.99E+05 8.289E+03 Tank contents are released over 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months rather than 5 minutes.

Essentially all activity 99.9% is released over 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months at this reduced release rate.

Serial Number 11-025A Page 165 of 191 Table 3.8-2 Basic Data and Assumptions for WGDT Parameter or Assumption CLB Value Proposed Value Reason for Change EAB X/Q (sec/m3) 0 - 2 hr 2.232E-04 1.76E-04 New PAVAND X/Q values (see Table 1.3-3 and Section 3.1.2)

LPZ X/Q (sec/m3)

Period LPZ 0 - 2 hr 3.977E-05 2 - 24 hr 4.100E-06 1 - 2 day 2.427E-06 2 - 30 day 4.473E-07 Period LPZ 0 - 8 hr 3.36E-05 8 - 24 hr 2.37E-05 1 - 4 day 1.12E-05 4 - 30 day 3.94E-06 New PAVAND X/Q values (see Table 1.3-3 and Section 3.1.2)

Control Room Control Room Volume (ft3) 127,600 No Change Normal Ventilation Unfiltered Makeup Air Flow (scfm) 2,750 No Change Filtered Recirculation Air Flow (scfm) 2,250 No Change Control Room Post Accident Recirculation system (CRPARS) Ventilation (min) 0.5 NA Control room post accident recirculation is not credited to maximize control room dose Control Room Isolation (min) 0.5 30 The new proposed analysis assumes a bounding value of control room isolation time that will

Serial Number 11-025A Page 166 of 191 Table 3.8-2 Basic Data and Assumptions for WGDT Parameter or Assumption CLB Value Proposed Value Reason for Change maximize control room dose.

Analyses performed assuming NO isolation produce control room consequences that are less than the proposed case with 30 minute isolation.

CRPARS Filter Efficiency

(%)

Elemental Organic Particulate 90 (includes safety factor of 2) 90 (includes safety factor of 2) 99 NA Control room post accident recirculation is not credited to maximize control room dose.

Control Room Unfiltered Inleakage (cfm) 0 200 Low inleakage is assumed to maximize control room dose, but the grossly conservative assumption of no unfiltered inleakage was eliminated.

Sensitivity cases show that higher inleakage will result in lower predicted dose.

Maximum (ASTM) E741 tracer gas test = 447+/-51 cfm (Ref. 20)

Minimum (ASTM) E741 tracer gas test = 409 +/-29 cfm (Ref. 20)

Serial Number 11-025A Page 167 of 191 Table 3.8-2 Basic Data and Assumptions for WGDT Parameter or Assumption CLB Value Proposed Value Reason for Change Control Room HVAC Parameters (cfm)

Unfiltered Inleakage Unfiltered Make-up Air Filtered Recirculation 0 - 0.5 m 0.5 m - 30 d 0

0 2750 0

0 2250 0 - 30 m 30 m - 30 d 0

200 2750 0

0 0

Unfiltered inleakage is not assumed until the control room is isolated at 30 minutes at which time inleakage is assumed at 200 cfm (discussed above).

Control Room X/Q (sec/m3) 0 - 8 h 2.93E-3 0 - 2 h 3.67E-3 NEW ARCON96 X/Q values The highest calculated 0-2 hour X/Q value from the Auxiliary Building release pathway to the control room intake is from the Auxiliary Building Stack Exhaust.

Serial Number 11-025A Page 168 of 191 3.8.6 WGDT Analysis Results The results of the design basis WGDT analysis are presented in Table 3.8-3. These results report the calculated dose for the worst 2-hour interval (EAB), and for the assumed 30-day duration of the event for the control room and the LPZ. The EAB and LPZ doses are calculated with RADTRAD and are compared with the applicable acceptance criteria specified in original licensing basis and Branch Technical Position 11-5, based on the earlier version of 10 CFR 20. Control Room dose is compared with the limits defined in General Design Criteria 19 (Reference 31) and applicable standards in RG 1.183.

Table 3.8-3 Dose Results for the WGDT Accident Location (rem)

Limits (rem)

EAB 0.1 (WB) 0.5 (WB)

LPZ 0.1 (WB) 0.5 (WB)

Control Room 0.4 (TEDE) 5 (TEDE)

The results in Table 3.8-3 represent the highest control room and offsite doses that would result from a WGDT accident using worst case scenario conditions. As discussed previously, the control room consequences above assume control room isolation and unfiltered inleakage assumptions that maximize control room dose.

Control room dose in an unisolated control room will actually be less than the value listed in Table 3.8-3.

Serial Number 11-025A Page 169 of 191 3.9 Volume Control Tank Rupture (VCT) Analysis This section describes the methods employed and results of the Volume Control Tank rupture (VCT) design basis radiological analysis. The analysis assumes a failure of the VCT system that results in the release of the contents of the tank and additional releases from letdown flow until the letdown path is isolated. Doses were calculated at the Exclusion Area Boundary (EAB), at the Low Population Zone (LPZ), and in the KPS Control Room. The methodology used to evaluate the control room and offsite doses resulting from a VCT accident included Standard Review Plan Branch Technical Position 11-5 (Reference 19), Regulatory Guide 1.24 (Reference 30), ARCON96-based control room atmospheric dispersion factors, PAVAND-based EAB and LPZ atmospheric dispersion factors, and Federal Guidance Reports (FGR) No. 11 and 12 dose conversion factors.

The current VCT analysis credits the control room post accident recirculation system (CRPARS) in the determination of control room dose. New analyses have been performed to demonstrate that the CRPARS is not required to maintain control room dose within acceptable limits.

3.9.1 VCT Scenario Description The Volume Control Tank rupture (VCT) accident is defined as an unexpected and uncontrolled release to the atmosphere of the radioactive noble gas and halogen activity contained in the VCT and additional releases from radioactivity contained in letdown flow until isolated. Rupture of the volume control tank is assumed to release all the contained noble gases and one percent of the halogen inventory of the tank plus that amount contained in the 88-gpm flow from the demineralizers, which would continue for up to five minutes before isolation.

Serial Number 11-025A Page 170 of 191 3.9.2 VCT Source Term Definition 3.9.2.1 Activities The activities assumed in the VCT analysis and shown in Tables 3.9-1, 3.9-2 and 3.9-3, were previously calculated for the stretch power uprate and approved in Amendment No. 172, (Reference 11). They represent the maximum activity of noble gases and halogens accumulated within the VCT and available for release. The inventory of gases in the tank is based on continuous operation with one percent fuel defects and without any purge of the gas space. The inventory of iodine in the tank is based on operation of the plant with one percent fuel defects and with 90 percent of the iodine removed by the letdown demineralizer. The maximum VCT inventory for each nuclide is given in Table 3.9-1. The concentration of noble gases in the letdown flow is listed in Table 3.9-2. In addition, a pre-accident iodine spike is assumed, although not required. The current assumption of a spike of 60 Ci/gm dose equivalent I-131 (DEI) is being maintained even though the limit is being reduced to 10 Ci/gm DEI in this license amendment request. The iodine concentration in the letdown flow is listed in Table 3.9-3.

The activities being assumed for the VCT rupture are extremely conservative compared to actual activity which would exist in the volume control tank and letdown line, based on revised reactor coolant activity limits. New Technical Specification requirements will limit reactor coolant activity (see Sections 2.3 and 3.4.2) to a small fraction of current requirements. The source terms represented in Tables 3.9-1, 3.9-2 and 3.9-3 do not include the associated reductions that would be caused by operating at reduced RCS limits.

3.9.2.2 Source Term Multiplier The current licensing basis analysis for the VCT rupture includes a multiplier of 1.1 that is applied to the resulting calculated dose to allow for minor variations in fuel designs (e.g., core mass of 49.1 MTU +/- 10%, enrichment of 4.5 w/o +/- 10%, and cycle length of 493.6 EFPD +/- 5%). This 10% increase considers allowances for letdown flow variation (5.5%), VCT water volume variation (2.5%) and fuel management variation

Serial Number 11-025A Page 171 of 191 (2%). Each of these allowances are individually conservative in their application.

Together, they provide approximately a factor of two conservatism above the expected increase necessary to account for such variations.

To make the revised VCT analysis consistent with the assumptions applied to other analyses in this report that provide allowance for minor variations in fuel design, the variation in cycle length was increased from 493.6 EFPD +/- 5% to 493.6 EFPD +/-

10%. For noble gases and iodines, this variation has the effect of doubling the conservative fuel management variation from 2% to 4% based on sensitivity studies performed by Westinghouse. Therefore, the revised source term multiplier will increase from 1.1 to 1.12.

3.9.3 VCT Release Transport The release scenario assumes the failure of the volume control tank or piping, releasing activity into the Auxiliary building. No credit is taken for building volume dilution. As a result of the accident, all of the noble gas in the tank and one percent of the iodine in the tank liquid is assumed to be released to the atmosphere over a period of 5 minutes.

After event initiation, letdown flow to the volume control tank continues at the maximum flow rate of 88 gpm (maximum letdown flow plus 10-percent uncertainty) for 30 minutes when the letdown line is assumed to be isolated. The release is modeled using the Auxiliary Building Stack Exhaust as the release point to maximize the control room dose. The effluent resulting from the postulated event is assumed to release to the environment without continuous effluent radiation monitoring to automatically isolate and/or terminate the effluent release.

Serial Number 11-025A Page 172 of 191 Table 3.9-1 Volume Control Tank Activity (Ci)

Nuclide Activity in VCT (Ci)

Kr-85m 6.29E+01 Kr-85 7.35E+02 Kr-87 1.64E+01 Kr-88 8.85E+01 Xe-131m 2.07E+02 Xe-133m 2.21E+02 Xe-133 1.62E+04 Xe-135m 2.79E+01 Xe-135 4.52E+02 Xe-138 1.94E+00 I-131 8.69E-01 I-132 8.85E-01 I-133 1.30E+00 I-134 1.79E-01 I-135 7.09E-01

Serial Number 11-025A Page 173 of 191 Table 3.9-2 Letdown Flow Noble Gas Concentration (Ci/gm)

Nuclide Concentration (Ci/gm)

Kr-85m 1.73 Kr-85 8.60 Kr-87 1.13 Kr-88 3.28 Xe-131m 3.04 Xe-133m 3.44 Xe-133 242 Xe-135m 0.501 Xe-135 8.69 Xe-138 0.628 Table 3.9-3 Pre-Accident Iodine Spike Concentration based on 60 Ci/gm DEI Nuclide Concentration (Ci/gm)

I-131 46.8 I-132 47.6 I-133 69.8 I-134 9.7 I-135 38.2

Serial Number 11-025A Page 174 of 191 3.9.4 VCT Atmospheric Dispersion Factors 3.9.4.1 VCT Control Room /Qs As described in Section 3.1, the onsite atmospheric dispersion factors were calculated using the ARCON96 code (Reference 5) and guidance from Regulatory Guide 1.194 (Reference 6). The VCT Control Room X/Qs listed in Table 1.3-4 were calculated for the following applicable KPS source point:

Auxiliary Building Stack Exhaust The control room X/Q determined for the Auxiliary Building Stack Exhaust to the Control Room Intake represents the highest value applicable to any source to receptor combination for the VCT accident.

3.9.4.2 VCT Offsite (EAB & LPZ) /Qs As described in Section 3.1, the Exclusion Area Boundary (EAB) and Low Population Zone (LPZ) atmospheric dispersion factors were revised and are listed in Table 1.3-3.

These offsite atmospheric dispersion factors were generated using the PAVAND code (Reference 7) and guidance from Regulatory Guide 1.145 (Reference 8). The EAB and LPZ X/Q values were modeled using a ground-level release without credit for building wake to determine a conservative short-term diffusion estimate (X/Q).

3.9.5 VCT Analysis Assumptions and Key Parameters 3.9.5.1 Method of Analysis The RADTRAD-NAI code (Reference 3) is used to calculate the radiological consequences from airborne releases resulting from a VCT rupture at Kewaunee Power Station (KPS) to the EAB, LPZ, and Control Room.

The total contents of the VCT are assumed to release within 5 minutes. Assuming a removal rate of 82.9/hr will release essentially all (99.9%) of the tank contents over a five minute period. This equates to a release rate, for use in RADTRAD, of 1.99E+05%/day. The letdown release rate of 88 gallons per minute assumes 1% of the

Serial Number 11-025A Page 175 of 191 iodines after passing through the demineralizer in addition to a VCT tank DF of 10. This was modeled in RADTRAD as a filter with a filter efficiency set to 99.9% to mimic the iodine removal.

There are aspects of the VCT analysis that required multiple RADTRAD models.

Conservative combinations of control room ventilation rates (filtered recirculation and unfiltered inleakage) and control room isolation times (varied from quick isolation to delayed isolation to no isolation) were modeled in order to maximize control room dose and prove the control room post accident recirculation system (CRPARS) is not needed to maintain control room dose within acceptable limits (see Section 3.9.5.3).

A schematic shown in Figure 3.9-1 provides an overall picture of the VCT release to the environment modeled in RADTRAD.

Figure 3.9-1 VCT Radioactive Release Schematic Auxiliary Building VCT release of 1.99E+5%/day 5 minute release 1% of iodine released Flow = 88 gpm (3.33E+5 gm/min)

Letdown flow (after demin) 30 minute release Table 3.9-2 Iodine filter set to 99.9%

Table 3.9-3 3.9.5.2 Basic Data & Assumptions for VCT Changes have been made to the VCT analysis. Table 3.9-4 provides a complete list of inputs and assumptions used to reanalyze the KPS VCT.

VCT Table 3.9-1 1% failed fuel VCT Curies Environment Environment

Serial Number 11-025A Page 176 of 191 3.9.5.3 Assumptions to Maximize Control Room Dose Control room dose was calculated with a combination of control room assumptions that maximize control room dose for the two release pathways modeled for a VCT rupture.

One pathway is the tank rupture and near instantaneous release of tank radioactive contents within 5 minutes. The other pathway is the continual release of activity contained in letdown flow that will persist into the VCT and out of the ruptured tank until such time that letdown is isolated. Each pathway was evaluated for a condition of no control room isolation and delayed control room isolation. In both instances, delayed isolation produces higher control room dose. Results from both pathways were summed. The combination of assumptions for control room isolation and unfiltered inleakage rate that maximize control room dose were determined. The dose results, presented in Table 3.9-5, bound the condition of no control room isolation.

For the VCT rupture pathway, the set of control room assumptions that produced the highest control room dose was a 2.5 minute delayed isolation of the control room in conjunction with low unfiltered inleakage. For the letdown line release pathway that persists for 30 minutes until letdown is isolated, the set of control room assumptions that will maximize control dose was a 30 minute isolation of the control room in conjunction with low unfiltered inleakage. This combination of assumptions maximize the dose to inhabitants of the control room. Intake of the radioactive material at the maximum control room ventilation intake rate with delayed isolation of the control room and reduced intake/outflow will trap the radioactivity within the control room and maximize the exposure to the occupants. Time sensitivity runs determined that 2.5-minute isolation maximized the 5-minute VCT rupture pathway scenario and 30-minute isolation maximized the 30-minute letdown pathway scenario. Assuming no control room isolation or isolation prior to or after the times listed, produce lower dose consequences.

Unfiltered inleakage rate sensitivity runs showed that control room dose is maximized by assuming a low unfiltered inleakage rate. An unfiltered inleakage rate of 200 cfm was determined to maximize control room dose. This rate is lower than one half of the unfiltered inleakage air flow of 409 +/- 29 cfm measured by the American Society for

Serial Number 11-025A Page 177 of 191 Testing and Materials (ASTM) E741 (tracer gas) leakage test conducted in December 2004 (Reference 20). Unfiltered inleakage rates greater than 200 cfm produce lower control room dose due to the associated purge effect of the inflow.

The design assumptions for the VCT rupture analysis that maximize dose and demonstrate that the control room post accident recirculation system is not needed are listed in Table 3.9-4.

Serial Number 11-025A Page 178 of 191 Table 3.9-4 Basic Data and Assumptions for VCT Parameter or Assumption CLB Value Proposed Value Reason for Change Source Term VCT Radiation Source (Curies)

Table 3.9-1 No Change Letdown Line RCS Noble Gas Concentration (Ci/gm)

Table 3.9-2 No Change Letdown Line Pre-Accident Iodine Spike Concentration (Ci/gm)

Table 3.9-3 [conservatively based on spike of 60 Ci/gm DEI]

No Change Iodine Release from VCT and Letdown Line (%)

1 No Change Source Term Multiplier 1.1 1.12 Source term adjustment factor to allow for fuel management variations. Previous multiplier provided by Westinghouse allowed variation in cycle length of 493.6 +/-

5% EFPD. The new higher multiplier accounts for a larger fuel management variation, similar to that required in the RSAC of 493.6 EFPD

+/- 10% (see Section 3.9.2.2).

RCS Coolant Activity

(% failed fuel) 1 No Change Core Activity Table 3.2-3 No Change

Serial Number 11-025A Page 179 of 191 Table 3.9-4 Basic Data and Assumptions for VCT Parameter or Assumption CLB Value Proposed Value Reason for Change Demineralizer Iodine DF for Letdown Flow prior to VCT 10 No Change VCT Parameters Release Duration (min)

VCT Letdown 5

5 No Change 30 Release Rate (%/day) 1.99E+05 No Change EAB X/Q (sec/m3) 0 - 2 hr 2.232E-04 1.76E-04 New PAVAND X/Q values (see Table 1.3-3 and Section 3.1.2)

LPZ X/Q (sec/m3)

Period LPZ 0 - 2 hr 3.977E-05 2 - 24 hr 4.100E-06 1 - 2 day 2.427E-06 2 - 30 day 4.473E-07 Period LPZ 0 - 8 hr 3.36E-05 8 - 24 hr 2.37E-05 1 - 4 day 1.12E-05 4 - 30 day 3.94E-06 New PAVAND X/Q values (see Table 1.3-3 and Section 3.1.2)

Letdown Flow (gpm) 88 No Change RCS Mass (grams) 1.192E+08 No Change Control Room Control Room Volume (ft3) 127,600 No Change

Serial Number 11-025A Page 180 of 191 Table 3.9-4 Basic Data and Assumptions for VCT Parameter or Assumption CLB Value Proposed Value Reason for Change Normal Ventilation Unfiltered Makeup Air Flow (scfm) 2,750 No Change Filtered Recirculation Air Flow (scfm) 2,250 No Change Control Room Post Accident Recirculation system (CRPARS) Ventilation (min) 0.5 NA Control room post accident recirculation is not credited to maximize control room dose.

Control Room Isolation (min)

VCT Letdown 0.5 0.5 2.5 30 The new Proposed analysis assumes a bounding value of control room isolation times that will maximize each pathway control room dose. Analyses performed assuming NO isolation produce control room consequences that are lower.

CRPARS Filter Efficiency

(%)

Elemental Organic Particulate 90 (includes safety factor of 2) 90 (includes safety factor of 2) 99 NA Control room post accident recirculation is not credited to maximize control room dose.

Control Room X/Q (sec/m3) 0 - 8 h 2.93E-03 0 - 2 h 3.67E-03 NEW ARCON96 X/Q values The highest calculated 0-2 hour X/Q value of from any possible release pathway from the Auxiliary Building to the control room intake is from the Auxiliary Building Stack

Serial Number 11-025A Page 181 of 191 Table 3.9-4 Basic Data and Assumptions for VCT Parameter or Assumption CLB Value Proposed Value Reason for Change Exhaust.

EAB and LPZ X/Q (sec/m3)

EAB 2.232E-04 LPZ (0-2 hr) 3.977E-05 (2-24 hr) 4.100E-06 (1-2 day) 2.427E-06

>2 day 4.473E-07 Table 1.3-3 NEW PAVAND X/Q values Control Room Unfiltered Inleakage (cfm) 0 200 The grossly conservative assumption of no unfiltered inleakage was raised but still remains lower than measured inleakage. Low inleakage is assumed to maximize control room dose. Sensitivity cases show that higher inleakage will result in lower predicted dose.

Maximum (ASTM) E741 tracer gas test = 447+/-51 cfm (Ref. 20)

Serial Number 11-025A Page 182 of 191 Table 3.9-4 Basic Data and Assumptions for VCT Parameter or Assumption CLB Value Proposed Value Reason for Change Control Room HVAC Parameters (cfm)

VCT Unfiltered Inleakage Unfiltered Make-up Air Filtered Recirculation Letdown Unfiltered Inleakage Unfiltered Make-up Air Filtered Recirculation 0 - 0.5 m 0.5 m - 30 d 0

0 2750 0

0 2250 0 - 0.5 m 0.5 m - 30 d 0

0 2750 0

0 2250 0 - 2.5 m 2.5 m - 30 d 0

200 2750 0

0 0

0 - 0.5 hr 0.5 hr - 30 d 0

200 2750 0

0 0

Unfiltered inleakage of 200 cfm is not assumed until the control room is isolated.

A conservative combination of control room isolation times and inleakage assumptions were used to maximize control room dose.

Analyses performed assuming NO isolation produce control room consequences that are lower.

Serial Number 11-025A Page 183 of 191 3.9.6 VCT Analysis Results The results of the design basis VCT analysis are presented in Table 3.9-5. These results report the calculated dose for the worst 2-hour interval (EAB), and for the assumed 30-day duration of the event for the control room and the LPZ. The EAB and LPZ doses are calculated with RADTRAD and are compared with the applicable acceptance criteria specified in original licensing basis and Branch Technical Position 11-5, based on the earlier version of 10 CFR 20. Control Room dose is compared with the limit specified in General Design Criteria 19 (Reference 31) and applicable standards in RG 1.183.

Table 3.9-5 Dose Results for the VCT Accident Location (rem)

Limits (rem)

EAB 0.1 (WB) 0.5 (WB)

LPZ 0.1 (WB) 0.5 (WB)

Control Room 0.6 (TEDE) 5 (TEDE)

The results in Table 3.9-5 represent the highest control room and offsite doses that would result from a VCT accident using worst case scenario conditions. As discussed previously, the control room consequences above were calculated using control room isolation and unfiltered inleakage assumption combinations that will maximize control room dose. Control room dose in an unisolated control room would be less than the value listed in Table 3.9-5.

Serial Number 11-025A Page 184 of 191 4.0 ADDITIONAL DESIGN BASIS CONSIDERATIONS In addition to the explicit evaluation of radiological consequences that had direct impact from the changes associated with this request, other areas of plant design were also considered for potential impacts. The evaluation of these additional design areas is documented below.

4.1 Risk Impact of Proposed Changes The proposed changes associated with implementation of the revised design basis radiological analyses for Kewaunee Power Station have been considered for their risk effects. A discussion of these considerations is presented below.

The proposed changes are presented here for convenience; these changes are described in report Section 2:

a. Revised Meteorological X/Q Values for Off-site and Control Room Receptors

b. Use of the RADTRAD-NAI Code to analyze Dose Consequences

c. Reduction in Maximum RCS Coolant Activity Limits

d. Reduction in SG Secondary Coolant Activity Limit

e. Isolation of the Control Room prior to moving Recently Irradiated Fuel

f. Refueling Operation Requirements to allow Open Containment Penetrations

g. Elimination of R-23 Credit for Control Room Isolation

h. Revise Technical Specification Definition of Dose Equivalent I-131

i. Changes in Design and License Basis Assumptions Item a The change in X/Q values has a direct effect on calculated dose consequences. The new values were calculated pursuant to the guidance of Regulatory Guides 1.145 and 1.194, respectively. Their use in design basis analyses assure that the resulting consequences contain sufficient conservatism due to atmospheric dispersion such that the value is not

Serial Number 11-025A Page 185 of 191 exceeded by more than 5.0 percent of the time. This change has no impact upon plant risk.

Item b RADTRAD-NAI designed after the ITSC version of RADTRAD developed for the NRC, has been previously found to be acceptable for use in dose calculations. Its use has no impact upon plant risk.

Item c The reduction in maximum allowed RCS coolant activity in Technical Specifications will cause a commensurate reduction in potential dose consequences as a result of RCS releases. Reducing RCS concentration has no impact upon plant risk.

Item d The reduction in maximum allowed secondary side activity will cause a commensurate reduction in potential dose consequences as a result of secondary side releases. Reducing secondary side concentration has no impact upon plant risk.

Item e Control room isolation will be required prior to moving recently irradiated fuel. This measure was necessary to eliminate credit for R-23 and maintain control room dose within limits. Having the control room isolated does not impact plant risk.

Item f Allowing containment penetrations to be open during movement of recently irradiated fuel has been shown to result in acceptable off-site consequences. The design analysis assumes the containment remains open for the entire 2-hour duration of the fuel handling event. Being under Technical Specification required Administrative Control, the ability and likelihood for closure of open containment penetrations, in the event of an accident, is increased. Closure of penetrations is an additional defense-in-depth, not credited - but available if conditions warrant. Allowing penetrations to be open during refueling provides flexibility in outage

Serial Number 11-025A Page 186 of 191 scheduling and additional comfort to workers. Open penetrations have no impact upon plant risk.

Item g The elimination of control room isolation credit from the control room inlet monitor R-23 removes reliance for a safety function performed by instrumentation that is not redundant, not safety grade, and provides incomplete isolation of the control room. Credit for R-23 currently exist for the FHA and LRA. With the requirement to require control room isolation prior to movement of recently irradiated fuel, plant risk is not impacted and the FHA control room consequences are acceptable. Crediting Operator action to isolate the control room within 1-hour after LR event will result in acceptable dose consequences. Current license basis credits Operator action within 45 minutes of a LRA if R-23 fails to perform its safety function. Credit for Operator action has been extended to 1-hour, reducing timing burden on the Operator. Extending the allowed time to isolate control room following a LRA does not impact plant risk. No longer crediting R-23 for control room isolation is acceptable and removes future vulnerability by reliance on non-safety grade instrumentation to perform a safety function.

Item h Changing the Technical Specification definition of DEI allows the use of dose conversion factors from FGR 11. These dose conversion factors have been previously found to be acceptable for use in dose calculations.

This change has no impact upon plant risk from severe accident scenarios.

Item i The changes in design and license basis assumptions have been evaluated for the full spectrum of USAR Chapter 14 design basis analyses. The changes proposed in congregate form have been demonstrated to result in acceptable off-site and control room dose consequences. All assumptions have been validated and are presented

Serial Number 11-025A Page 187 of 191 for approval with associated discussions on methods and inputs. The changes are deemed safe and do not pose an impact on plant risk.

The revised assessments of the radiological consequences due to design basis accidents listed in the KPS USAR, using the AST methodology and proposed assumptions and inputs, conclude that the EAB, LPZ, and Control Room doses are within the limits of 10 CFR 50.67 and within the limits of Regulatory Guide 1.183. The results of this proposed amendment demonstrate that there will be no adverse impact on public health and safety.

4.2 Impact Upon the Emergency Plan This proposed revision to Technical Specifications and USAR design basis analyses will replace the existing radiological licensing basis upon approval. The current Emergency Action Levels (EAL) for KPS implement the NEI 99-01 Rev. 4 (Reference 29) guidance.

EAL limits (SU 4.1 and SU 4.2 for hot conditions and CU 5.1 for cold shutdown) apply criteria that relate reactor coolant sample activity and associated radiation monitor readings to provide indication of fuel clad integrity. These limits, if exceeded, are considered to be a potential degradation in the level of safety of the plant and a potential precursor of more serious problems. The current limits are tied to Technical Specification 3.4.16 limits on RCS activity and spikes. With the proposed changes to TS 3.4.16 to reduce RCS activity limits, corresponding changes to these three EAL limits will be necessary to maintain the same level of effectiveness and maintain the same technical basis. SU 4.1 and CU 5.1 letdown radiation monitor (R-9) limits based on an RCS concentration of 1.0 Ci/gm DE I-131 will need to be reduced by a factor of ten to correspond to the proposed RCS activity limit reduction. This new limit remains sufficiently above normal background readings on R-9 to provide indication of a degraded fuel condition. Likewise, SU 4.2 RCS activity limits which are based on TS 3.4.16 will need to be revised to correspond with the new proposed technical specification limits for RCS activity.

Serial Number 11-025A Page 188 of 191 Other than RCS activity limit reductions, design basis source terms were unaffected by this license amendment request. In addition, revised design basis X/Q dispersion factors are not used by the emergency plan. Therefore, beyond the above identified change to the EALs, existing emergency plan procedures and dose assessment tools and models are unaffected by the changes proposed in this request.

5.0 Conclusions The proposed changes in Technical Specifications, design assumptions, and offsite and control room X/Qs have been incorporated into the reanalysis of radiological effects from eight key accidents for KPS. The analysis results from the reanalyzed events meet all of the acceptance criteria as specified in 10 CFR 50.67, RG 1.183, and BTP 11-5.

Serial Number 11-025A Page 189 of 191 6.0 References

1.

Regulatory Guide 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, USNRC, Office of Nuclear Regulatory Research, July 2000.

2.

10 CFR 50.67, Accident Source Term

3.

Software - RADTRAD-NAI Version 1.1a(QA), Numerical Applications Inc.

4.

NUREG/CR-6604, "RADTRAD: A Simplified Model for Radionuclide Transport and Removal and Dose Estimation, USNRC, June 1997, S.L.

Humphreys et al.

5.

NUREG/CR-6331, Rev. 1, Atmospheric Relative Concentrations in Building Wakes, ARCON96, USNRC, 1997.

6.

Regulatory Guide 1.194, Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants, U.S.

Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, June 2003

7.

Software - PAVAND Version 1-00, Atmospheric Dispersion Model.

8.

Regulatory Guide 1.145, Revision 01, Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants, February 1983.

9.

Letter K-88-032, K.E. Perkins, USNRC to D.C. Hintz, WPSC; Application of Leak-Before-Break Technology as a Basis for Kewaunee Nuclear Power Plant Steam Generator Snubber Reduction, 2/16/88.

10. Letter from John Lamb (NRC) to Tom Coutu (NMC) transmitting Issuance of Alternate Source Term Amendment # 166 (TAC No. MB4596) and Safety Evaluation Report dated March 17, 2003.

11. Letter from John Lamb (NRC) to Tom Coutu (NMC) transmitting Issuance of Stretch Power Uprate Amendment #172 (TAC No. MB9031) and Safety Evaluation Report dated February 27, 2004.

12. License Amendment Request 211, Radiological Accident Analysis and Associated Technical Specifications Change, dated January 30, 2006, (ML060540217).

Serial Number 11-025A Page 190 of 191

13. RAI Response Regarding License Amendment Request - 211 Radiological Accident Analysis and Associated Technical Specifications Change, dated January 23, 2007 (ML070240543).

14. Letter from R. F. Kuntz (NRC) to D. A. Christian (Dominion) transmitting issuance of Radiological Accident Analysis Amendment #190 (TAC No.

MC9715) and Safety Evaluation Report dated March 8, 2007.

15. Federal Guidance Report No. 11, Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion, EPA 520/1-88-020, Environment Protection Agency, 1988.

16. Federal Guidance Report No. 12, External Exposures to Radionuclides in Air, Water and Soil, EPA 420-r-93-081, Environmental Protection Agency, 1993.

17. Not Used.

18. Regulatory Guide 1.23, Revision 1, Meteorological Monitoring Programs for Nuclear Power Plants, March 2007.

19. NRC Branch Technical Position ETSB 11-5, Postulated Radioactive Releases due to a Waste Gas System Leak or Failure, Rev 0, July 1981.

20. Control Room Tracer Gas Test Report entitled, "Control Room Habitability Tracer Gas Leak Testing at the Kewaunee Nuclear Plant," dated January 27, 2005.

21. NUREG-0800, Standard Review Plan, Section 6.4, Control Room Habitability System, Revision 2, July 1981.

22. NUREG-0800, "Standard Review Plan, Section 6.5.2, "Containment Spray as a Fission Product Cleanup System, U.S. Nuclear Regulatory Commission, Revision 2, December 1988.

23. NUREG/CR-5966, A Simplified Model of Aerosol Removal by Containment Sprays, June 1993.

24. KPS Drawing M-358, Revison L, Reactor Building Piping - Internal Containment Spray, August 19, 2008.

25. NUREG/CR-6189, A Simplified Model of Aerosol Removal by Natural Processes in Reactor Containments, July 1996.

26. BNP-100, Iodine Removal from Containment Atmospheres by Boric Acid Spray, July 1970.

Serial Number 11-025A Page 191 of 191

27. WCAP-7828, Radiological Consequences of a Fuel Handling Accident, December 1997.

28. NSAL-00-004, Westinghouse Nuclear Safety Advisory Letter dated March 7, 2000, Non-conservatisms in Iodine Spiking Calculations.

29. NEI 99-01, Revision 4, Methodology for Development of Emergency Action Levels, January 2003.

30. Regulatory Guide 1.24, Assumptions Used for Evaluating the Potential Radiological Consequences of a Pressurized Water Reactor Radioactive Gas Storage Tank Failure (Safety Guide 24), March 1972.

31. 10 CFR 50, Appendix A, General Design Criteria for Nuclear Power Plants, Criterion 19 - Control Room (GDC 19).

32. Procedure GNP-05.16.06, Validation of Time Dependent Operator Actions, Revision 6.

33. RAI Response Regarding Millstone Unit 3 Stretch Power Uprate License Amendment Request Response to Question AADB-07-0107, dated January 18, 2008 (ML080280375).

34. License Submittal, North Anna Power Station Units 1 and 2, Proposed Technical Specification Change and Supporting Safety Analyses Revisions to Address Generic Safety Issue 191, dated October 3, 2006, (ML062850195).

v t e Letter Sequence Request
TAC:ME7110, Control Room Habitability, Administratively Control Containment Penetrations, Deletion of E BAR Definition and Revision to RCS Specific Activity Tech Spec, Revise Containment Requirements During Handling Irradiated Fuel and Core Alterations (Open)
Initiation Request, Request, Request, Request, Request, Request, Request, Request, Request Acceptance, Acceptance, Acceptance Administration Questions 15 May 2012 15 May 2012 1 December 2011 9 January 2012 13 January 2012 30 January 2012 6 March 2012 13 April 2012 10 October 2012 Answers 8 November 2011 25 January 2012 27 February 2012 12 March 2012 30 April 2012 6 June 2012 3 August 2012 9 August 2012 Results Approval, Approval Other: ML11290A148, ML12066A008, ML12166A006, ML12193A684
MONTHYEAR2012-03-12 [Table View]

ML11252A656

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