ML063110185

Attachment 5 - Calculation No. H-1-ZZ-MDC-1880, Revision 2IR0, Post-LOCA Eab, LPZ and CR Doses.

ML063110185
Person / Time
Site:	Hope Creek
Issue date:	01/03/2003
From:	Drucker M, Morrison G, Gita Patel NUCORE, Public Service Enterprise Group
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LCR H05-01, Rev. 1, LR-N06-0418 H-1-ZZ-MDC-1880, Rev 2IRO
Download: ML063110185 (85)

Text

Attachment 5 LR-N06-0418 LCR HOS-01, Rev. I Calculation No. H-1-ZZ-MDC-1880, Revision 21R0 Post-LOCA EAB, LPZ and CR Doses

NC.DE-AP.ZZ-0002(Q)

CALC NO.: H-1-ZZ-MDC-1880 , CALCULATION COVER SHEET Page 1 of 84 REVISION: A-URr" -- X C5b"-Zb CALC. TITLE: I Post-LOCA EAB, LPZ, and CR Doses

1. SHTS (CALC):. 84 # ATT I # SHTS: 13/3 # IDV/50.59 SHTS: 14/3 # TOTAL SHTS: 94 CHECK ONE:

O1 FINAL JR INTERIM (Proposed Plant Change) El FINAL (Future Confirmation Req'd) E VOID SALEM OR HOPE CREEK: E Q - LIST 0 IMPORTANT TO SAFETY [I NON-SAFETY RELATED HOPE CREEK ONLY: 00Q [30s '-Qsh [IF [--R 0 STATION PROCEDURES IMPACTED, IF SO CONTACT RELIABILITY ENGINEER El CDs INCORPORATED (IF ANY):

DESCRIPTION OF CALCULATION REVISION (IF APPL.):

Revision 1 of this calculation includes the deletion of FRVS recirculation charcoal filtration and FRVS vent heater, the reduction of CR unfiltered inleakage from 900 cfm to 350 cfm and ESF leakage from 10 gpm to 1 gpm, and the increase of core thermal power to 4,031 MWt. The 10CFR50.59 evaluation for DCP 80048085 applies to this documentation (Ref.

10.49). The revision of Procedure HC.RA-AP.ZZ-005 I(Q), Rev. 1, will be tracked by a NUTS order by Licensing after receipt of the SER from the Staff. Revision bars are not used in this calculation due to extensive nature of changes.

PURPOSE:

The purpose of this calculation is to determine the EAB, LPZ, and control room doses for Hope Creek Generating Station (HCGS) due to the reduced CR unfiltered inleakage to 350 cfm, deletion of FRVS recirc charcoal filters, deletion of FRVS vent heater, reduction of ESF leakage from 10 gpm to I gpm, and increase of core thermal power to 4,031 MWt. The analysis is performed using the Alternate Source Term (AST), the guidance in the Regulatory Guide 1.183, and the TEDE dose criteria.

CONCLUSIONS:

The Section 8.0 results of this analysis indicate that the following changes can be implemented using the AST and guidance in the Regulatory Guide 1.183: control room unfiltered inleakage reduction to 350 cfm, deletion of FRVS recirc charcoal filters, deletion of FRVS vent heater, and increase of core thermal power to 4,031 MWt . Adherence to guidance in the RG 1.183, use of the specific values and limits contained in the technical specifications, and use of the as-built post-accident performance of safety grade ESF functions provide the assurance for sufficient safety margin, including a margin to account for analysis uncertainties in the proposed uses of an AST and the associated facility modifications and changes to procedures.

The V&V of RADTRAD3.02 code demonstrates that RADTRAD produces the identical results within +/- 2% margin of error compared to the HABITI.0 results.

Procedure HC.RA-AP.ZZ-005 1(Q), Rev. 1, "Leakage Reduction Program," should be revised to incorporate the new ESF leakage limit of 1 gpm established in this analysis.

Printed Name / Signature Date ORIGINATOR/COMPANY NAME: Gopal J. Patel/NUCORE 11/27/02 REVIEWER/COMPANY NAME: Mark Drucker/NUCORE ym_ !4 11/29/02 VERIFIER/COMPANY NAME: Mark Drucker/NUCORE W _ 4 11/29/02 PSEG SUPERVISOR APPROVAL: Gregory Morrison /PSEG Nular Common Revision 9

CALCULATION CONTINUATION SHEET SHEET 2 of 84 CALC. NO.: H-I-ZZ-MDC-1880

REFERENCE:

LCR H02-01 & DCP 80048085 O. Patel/NUCORE, ORIGINATOR, DATE REV: 11/27/02 A/*"-"

M. Drucker/NUCORE, REVIEWERIVERIFIER, DATE 11/29/02 REVISION HISTORY Revision Issue Date Revision Description 0IR0 5/7/01 Initial Issue.

OIRI 5/16/01 Revised due to incorporation of the preliminary plant-specific core inventory, which will be confirmed via Order No. 80028003. The CR inleakage value was reduced to 900 efin from 1000 cfm to offset the impact of preliminary core inventory on the CR dose.

01R2 8/01/01 Revised the aerosol removal rate in main steam piping, horizontal projected pipe surface area, and equation calculating the aerosol deposition.

Incorporated the revised X/Qs.

0 10/08/01 All interim revisions are incorporated. Isotopic activity released to environment is added. This is an original issue.

Removed the credit of FRVS recirculation charcoal filter efficiencies, I ~, reduced the FRVS vent charcoal filter efficiencies, control room unfiltered inleakage, and ESF leakage from 10 gpm to I gpm, and increased the core thermal power level by 11.9% to be consistent with a proposed power uprate.

JRevised to assess radiological impact of extended power uprate core

_ _ _ inventory.

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CALCULATION CONTINUATION SHEET SHEET 3 of 84 CALC. NO.: H-I-ZZ-MDC-1880

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LCR H02-01 & DCP 80048085 G. Patel/NUCORE, ORIGINATOR, DATE REV: 11/27/02 M. Drucker/NUCORE, REVIEWERNVERIFIER, DATE 11/29/02 PAGE REVISION INDEX PAGE REV PAGE REV PAGE REV 1 1 42 1 84 1 2 1 43 1 85 1 3 1 44 1 86 1 4 1 45 1 Attachment 14.1 I 5 1 46 1 Attachment 14.2 1 6 1 47 1 Attachment 14.3 1 7 i 48 1 8 1 49 1 9 I 50 1 10 1 51 1 11 1 52 1 12 1 53 1 13 1 54 1 14 1 55 1 15 1 56 1 16 I 57 1 17 1 58 1 18 I 59 1 19 1 60 1 20 1 61 1 21 1 62 1 22 1 63 1 23 1 64 1 24 1 65 1 25 1 66 1 26 1 67 1 27 1 68 1 28 1 70 1 29 1 71 1 30 1 72 1 31 I 73 1 32 I 74 1 33 I 75 1 34 I 76 1 35 I 77 1 36 1 78 1 37 1 79 1 38 1 80 1 39 1 81 1 !1 1 40 1 82 1 41 1 83 I 1

I Nuclear Common Nuclear Common Revision 99 I Revision

CALCULATION CONTINUATION SHEET SHEET 4 of 84 CALC. NO.: H-I-ZZ-MDC-1880

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LCR H02-01 & DCP 80048085 G. Patel/NUCORE, ORIGINATOR, DATE REV: 1127/02 1 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 1129/02 TABLE OF CONTENTS Section Sheet No.

Cover Sheet 1 Revision History 2 Page Revision Index 3 Table of Contents 4 1.0 Purpose 5 2.0 Scope 5 3.0 Analytical Approach 5 4.0 Assumptions 7 5.0 Design Inputs 11 6.0 Methodology 20 7.0 Calculations 28 8.0 Results Summary 38 9.0 Conclusions/Recommendations 40 10.0 References 41 11.0 Tables 47 12.0 Figures 77 13.0 Affected Documents 83 14.0 Attachments 83 I Nuclear Common Revision 9 1 Nula Cmo evso

1.0 PURPOSE The purpose of this calculation is to evaluate the Exclusion Area Boundary (EAB), Low Population Zone (LPZ), and Control Room (CR) Post-LOCA doses for Hope Creek Generating Station due to:

• The reduction of CR unfiltered inleakage from 900 cfm to 350 cfin. The actual measured CR unfiltered inleakage is 206 cfm including the measuring uncertainty,

• The deletion of FRVS recirculation charcoal filters,

• The reduction of FRVS vent charcoal filter efficiencies for elemental and organic iodine due to removal of the safety grade heater,

• The reduction of ESF leakage from 10 gpm to 1 gpm,

• The increase of core thermal power to 4,031 MWt The final results of the analyses are shown in Section 8.0 of this calculation. The doses are calculated using the Alternate Source Term (AST), Regulatory Guide (RG) 1.183 requirements, NRC sponsored RADTRAD3.02 computer code, and Total Effective Dose Equivalent (TEDE) dose methodology.

2.0 SCOPE The scope of this evaluation covers the anticipated dose consequences of a Post-LOCA scenario for the HCGS. This calculation is being performed in support of Design Change Packages (DCPs) 80048085 &

80048085. As part of this analysis, the following licensing basis post-LOCA release paths are analyzed:

1. Containment Leakage.

2. Engineered Safety Feature (ESF) Leakage.

3. Main Steam Isolation Valve (MSIV) Bypass Leakage.

3.0 ANALYTICAL APPROACH The deletion of FRVS recirculation charcoal filters and FRVS vent heater, reduction of ESF leakage, and increased core thermal power changed the key design input parameters previously used in the analysis. Therefore, the CR and site boundary doses are reanalyzed using the as-built design inputs.

The RADTRAD3.02 computer code (Refs. 10.2 & 10.48) is used to calculate EAB, LPZ, and CR doses based on the containment leakage, ESF leakage, and MSIV leakage release paths using as-built design inputs and assumptions and guidance in Regulatory Guide 1.183 (Ref. 10.1). The structures, systems, and components capable of performing their safety functions during and following a safe shutdown earthquake (SSE) are credited in the analysis.

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CALCULATION CONTINUATION SHEET SHEET 6 of 84 CALC. NO.: H-I-ZZ-MDC-1880

REFERENCE:

LCR H02-01 & DCP 80048085 G. Patel/NUCORE, ORIGINATOR, DATE REV: 11/27/02 1 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 11/29/02 The NRC Safety Evaluation related to Hope Creek Generating Station Technical Specification Amendment No. 134 identified that PSEG Nuclear used RADTRAD computer code, Version 3.02, in its dose calculation for submittal (Ref. 10.54); therefore, use of the RADTRAD has been accepted as a part of the Hope Creek design basis.

Reference 10.55 addresses the incremental doses resulting from removal of the reactor well shield plugs before cold shutdown, which has no impact on the radiological consequences analyzed in this calculation because the removal of shield plugs does not impact the post-LOCA release paths. The incremental doses at various offsite and onsite locations are insignificant (Ref. 10.55, pages 7 & 8).

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CALCULATION CONTINUATION SHEET SHEET 7 of 84 CALC. NO.: H-I-ZZ-MDC-1880

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LCR H02-01 & DCP 80048085 G. Patel/NUCORE, ORIGINATOR, DATE REV: 11/27/02 1 M. Drucker/NUCORE, REVIEWERNVERIFIER, DATE 11I29/02 4.0 ASSUMPTIONS The following assumptions used in evaluating the offsite and control room doses resulting from a Loss of Coolant Accident (LOCA) are based on the requirements in the Regulatory Guide 1.183 (Ref. 10.1).

These assumptions become the design inputs in Sections 5.3 through 5.7 and are incorporated in the analyses.

4.1 Source Term Assumptions Acceptable assumptions regarding core inventory and the release of radionuclides from the fuel are provided in Regulatory Positions (RGP) 3.1 through 3.4 of Reference 10.1 as follows:

4.2 Core Inventory The assumed inventory of fission products in the reactor core and available for release to the containment is based on the maximum power level of 4,031 MWt corresponding to current fuel enrichment and fuel burnup, which is 1.22 times the HCGS original licensed thermal power of 3,293 MWt including the 2% instrumentation uncertainty.

4.3 Release Fractions and Timing The core inventory release fractions, by radionuclide groups, for the gap release and early in-vessel damage for a Design Basis Accident (DBA) LOCA are listed in Design Input 5.3.1.5. These fractions are applied to the equilibrium core inventory described in Design Input 5.3.1.3 (Ref. 10.1, Tables 1 & 4).

4.4 Radionuclide Composition The elements in each radionuclide group to be considered in design basis analyses are shown in Design Input 5.3.1.4 (Ref. 10.1, RGP 3.4).

4.5 Chemical Form The suppression pool water pH is greater than 7 during and following a LOCA (10.43, page 11).

Consequently, the chemical forms of radioiodine released to the containment can be assumed to be 95%

cesium iodide (CsI), 4.85 percent elemental iodine, and 0.15 percent organic iodide (Ref. 10.1, RGP 3.5 and A.2). These are shown in Design Inputs 5.3.1.7. With the exception of elemental and organic iodine and noble gases, fission products are assumed to be in particulate form (Ref. 10.1, RGP 3.5 and A.2).

4.6 Assumptions on Activity Transport in Primary Containment 4.6.1 The radioactivity released from the fuel is assumed to mix instantaneously and homogeneously throughout the free air volume of the primary containment.

4.6.2 Reduction in airborne radioactivity in the containment by natural deposition within the containment is credited using the RADTRAD3.02 Powers model for aerosol removal coefficient with a 10-percentile probability (Ref. 10.1 RGP A.3.2 & Ref. 10.2 Section 2.2.2.1.2).

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CALCULATION CONTINUATION SHEET SHEET 8 of 84 CALC. NO.: H-I-ZZ-MDC-1880

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LCR H02-01 & DCP 80048085 G. Patel/NUCORE, ORIGINATOR, DATE REV: 11/27/02 1 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 11/29/02 4.6.3 The primary containment is assumed to leak at the allowable Technical Specification peak pressure leak rate for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (Ref. 10.1, RGP A.3.7). For HCGS, this leakage is reduced to 50% of its TS value after the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> based on the post-LOCA containment pressure (Ref. 10.15) as shown in design input 5.3.2.5.

4.6.4 The HCGS drywell and suppression chamber may be purged for up to 500 hrs per year (Ref.

10.6.18). Normally, the containment is purged at <25% power level before or during a drywell entry in an outage. Per RG 1.183, RGP A.7, the radiological consequences from post-LOCA primary containment purging as a combustible gas or pressure control measure should be analyzed. If the primary containment purging is required within 30 days of the LOCA, the results of this analysis should be combined with consequences postulated for other fission product release paths to determine the total calculated radiological consequences from the LOCA.

However, HCGS has a safety grade hydrogen recombination system to control the post-accident combustible gas (Ref. 10.41 & 10.42). The post-LOCA containment pressure is reduced to less than 31 psia within a few days (Ref 10.15). Containment purging is not required for the combustible gas or pressure control measure within 30 days of the LOCA. Therefore, the release from containment purging is not analyzed.

4.7 Offsite Dose Consequences The following assumptions are used in determining the TEDE for a maximum exposed individual at EAB and LPZ locations:

4.7.1 The offsite dose is determined in the TEDE, which is the sum of the committed effective dose equivalent (CEDE) from inhalation and the deep dose equivalent (DDE) from external exposure from all radionuclides that are significant with regard to dose consequences and the released radioactivity (Ref. 10.1, RGP 4.1.1, Ref 10.7). The RADTRAD3.02 computer code (Ref. 10.2) performs this summation to calculate the TEDE.

4.7.2 The offsite dose analysis is performed using the RADTRAD3.02 code (Ref. 10.2), which uses the Committed Effective Dose (CED) Conversion Factors for inhalation. (Ref. 10.1, RGP 4.1.2, Refs. 10.7 & 10.8).

4.7.3 Since RADTRAD3.02 calculates Deep Dose Equivalent (DDE) using whole body submergence in semi-infinite cloud with appropriate credit for attenuation by body tissue, the DDE can be assumed nominally equivalent to the effective dose equivalent (EDE) from external exposure.

Therefore, the code uses DDE in lieu of EDE in determining TEDE (Ref. 10.1, RGP 4.1.4, and Ref 10.8).

4.7.4 The maximum EAB TEDE for any two-hour period following the start of the radioactivity release is determined and used in determining compliance with the dose acceptance criteria in 10 CFR 50.67 (Ref. 10.1, RGP 4.1.5 & RGP 4.4, and Ref. 10.4).

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CALCULATION CONTINUATION SHEET SHEET 9 of 84 CALC. NO.: H-I-ZZ-MDC-1880

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LCR H02-01 & DCP 80048085 G. Patel/NUCORE, ORIGINATOR, DATE REV: 11/27/02 1 M. Drucker/NUCORE, REVIEWERNVERIFIER, DATE 11/29/02 EAB Dose Acceptance Criteria: 25 Rem TEDE (50.67(b)(2)(i))

4.7.5 TEDE is determined for the most limiting receptor at the outer boundary of the low population zone (LPZ) and is used in determining compliance with the dose criteria in 10 CFR 50.67 (Refs.

10.1, RGP 4.1.6 and RGP 4.4 & Ref. 10.4).

LPZ Dose Acceptance Criteria: 25 Rem TEDE (50.67(b)(2)(ii))

4.7.6 No correction is made for depletion of the effluent plume by deposition on the ground (Ref. 10.1, RGP 4.1.7).

4.7.7 The breathing rates used for persons at offsite locations is given in Reference 10.1, RGPs 4.1.3 &

4.4. These rates are incorporated in design inputs 5.7.2 & 5.7.4.

4.8 Control Room Dose Consequences The following guidance is used in determining the TEDE for maximum exposed individuals located in the control room:

4.8.1 The CR TEDE analysis considers the following sources of radiation that will cause exposure to control room personnel (Ref. 10.1, RGP 4.2.1). See applicable Design Inputs 5.6.1 through 5.6.13.

" Contamination of the control room atmosphere by the intake or infiltration of the radioactive material contained in the post-accident radioactive plume released from the facility (via CR air intake),

" Contamination of the control room atmosphere by the intake or infiltration of airborne radioactive material from areas and structures adjacent to the control room envelope (via CR unfiltered inleakage),

" Radiation shine from the external radioactive plume released from the facility (external airborne cloud),

• Radiation shine from radioactive material in the reactor containment (containment shine dose), and

" Radiation shine from radioactive material in systems and components inside or external to the control room envelope, e.g., radioactive material buildup in recirculation filters (CR filter shine dose).

4.8.2 The radioactivity releases and radiation levels used for the control room dose are determined using the same source term, transport, and release assumptions used for determining the exclusion area boundary (EAB) and the low population zone (LPZ) TEDE values (Ref. 10.1, RGP 4.2.2).

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CALCULATION CONTINUATION SHEET ISHEET 10 of 84 CALC. NO.: H-I-ZZ-MDC-1880

REFERENCE:

LCR H02-01 & DCP 80048085 G. Patel/NUCORE, ORIGINATOR, DATE REV: 11/27/02 1 M. Drucker/NUCORE, REVIEWERNERIFIER, DATE I It29/02 4.8.3 The occupancy and breathing rate of the maximum exposed individual present in the control room are incorporated in design inputs 5.6.12 & 5.6.13 (Ref. 10.1, RGP 4.2.6).

4.8.4 10 CFR 50.67 (Ref. 10.4) establishes the following radiological criterion for the control room.

This criterion is stated for evaluating reactor accidents of exceedingly low probability of occurrence and low risk of public exposure to radiation, e.g., a large-break LOCA (Ref. 10.1, RGP 4.4).

CR Dose Acceptance Criteria: 5 Rem TEDE (50.67(b)(2)(iii))

4.8.5 Credit for engineered safety features that mitigate airborne activity within the control room is taken for control room isolation/pressurization and intake & recirculation filtration (Ref. 10.1, RGP 4.2.4). The control room design is often optimized for the DBA LOCA and the protection afforded for other accident sequences may not be as advantageous. In most designs, control room isolation is actuated by engineered safety feature (ESF) signals or radiation monitors (RMs). In some cases, the ESF signal is effective only for selected accidents, placing reliance on the RMs.

Several aspects of RMs can delay the isolation, including the delay for activity to build up to concentrations equivalent to the alarm setpoint and the effects of different radionuclide accident isotopic mixes on monitor response. The CR emergency filtration system is conservatively assumed to be initiated at 30 minutes (Design Input 5.6.5) after a LOCA, after the CR normal supply fan has been tripped.

4.8.6 The CR unfiltered in leakage is conservatively assumed to be 500 cfln (Design Input 5.6.7) during the CREF transition period of 30 minutes after a LOCA. A conservative model would consider the normal ventilation mode for the transition period, which is of short duration (less than two minutes) until the control room envelop is fully pressurized following CREF initiation.

Such a model would result in total unfiltered inleakage of 6,600 Jft(3000 ft 3/min x 2 min x 1.1

[for 10% variation in flow] = 6,600 ft3). The conservative assumption of 500 cfmn unfiltered inleakage during the transition period would result in 15,000 ft (500 ft3/min x 30 min = 15,000 ft3) unfiltered air, which is 2 times higher.

4.8.7 No credits for KI pills or respirators are taken (Ref. 10.1, RGP 4.2.5).

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CALCULATION CONTINUATION SHEET SHEET 11 of 84 CALC. NO.: H-1-ZZ-MDC-1880

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LCR H02-01 & DCP 80048085 G. Patel/NUCORE, ORIGINATOR, DATE REV: 11/27/02 1 M. Drucker/NUCORE, REVIEWERNERIFIER, DATE 11/29/02 5.0 DESIGN INPUTS:

5.1 General Considerations 5.1.1 Applicability of Prior Licensing Basis The implementation of an AST is a significant change to the design basis of the facility and assumptions and design inputs used in the analyses. The characteristics of the AST and the revised TEDE dose calculation methodology may be incompatible with many of the analysis assumptions and methods currently used in the facility's design basis analyses. The HCGS plant specific design inputs and assumptions used in the TID-14844 analyses were assessed for their validity to represent the as-built condition of the plant and evaluated for their compatibility to meet the AST and TEDE methodology.

The analysis in this calculation ensures that analysis assumptions, design inputs, and methods are compatible with the requirements of the AST and the TEDE criteria.

5.1.2 Credit for Engineered Safety Features Credit is taken only for those accident mitigation features that are classified as safety-related, are required to be operable by technical specifications, are powered by emergency power sources, and are either automatically actuated or, in limited cases, have actuation requirements explicitly addressed in emergency operating procedures. The single active component failure modeled in this calculation is an

'A' or 'B' EDG failure concurrent with a loss of offsite power (LOP) resulting in the MSIV release at the ground level instead of released through the south plant vent (SPV). The consequences of an EDG failure is translated throughout the calculation by assuming that only four out of six FRVS recirculation filtration trains are available and one out of four inboard MSIV fails open. Assumptions regarding the occurrence and timing of a LOP are selected for the CREF system with the objective of maximizing the postulated radiological consequences.

5.1.3 Assignment of Numeric Input Values The numeric values that are chosen as inputs to analyses required by 10 CFR 50.67 are compatible to AST and TEDE dose criteria and selected with the objective of maximizing the postulated dose. As a conservative alternative, the limiting value applicable to each portion of the analysis is used in the evaluation of that portion. The use of containment, ESF, and MSIV leakage values higher than actually measured, use of 10% lower flow rates for the FRVS and CREFS recirculation systems, use of 10%

higher flow rate for FRVS vent, 30 minutes delay in the CREF initiation time, and use of ground release X/Qs demonstrate the inherent conservatisms in the plant design and post-accident response. Most of the design input parameter values used in the analysis are those specified in the Technical Specifications (Ref. 10.6).

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CALCULATION CONTINUATION SHEET SHEET 12 of 84 CALC. NO.: H-I-ZZ-MDC-1880

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LCR H02-01 & DCP 80048085 G. Patel/NUCORE, ORIGINATOR, DATE REV: 11/27/02 1 M. Drucker/NUCORE, REVIEWERNERIFIER, DATE 11/29/02 5.1.4 Meteorology Considerations Atmospheric dispersion factors (X/Qs) for the onsite release points such as the FRVS vent for containment and ESF leakage release path and turbine building louvers for MSIV leakage release path are re-developed (Ref. 10.5) using the NRC sponsored computer code ARCON96. The EAB and LPZ X/Qs are reconstituted using the HCGS plant specific meteorology and appropriate regulatory guidance (Ref. 10.32). The site boundary X/Qs reconstituted in Reference 10.32 were accepted by the staff in the previous licensing proceedings.

5.2 Accident-Specific Design Inputs/Assumptions The design inputs/assumptions utilized in the EAB, LPZ, and CR habitability analyses are listed in the following sections. The design inputs are compatible with the requirements of the AST and TEDE dose criteria and the assumptions are consistent with those identified in Regulatory Position 3 and Appendix A of RG 1.183 (Ref. 10.1). The design inputs and assumptions in the following sections represent the as-built design of the plant.

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Design Input Parameter Value Assigned Reference 5.3 Containment Leakage Model Parameters 5.3.1 Source Term 5.3.1.1 Thermal Power Level 4,031 MWt Section 7.9 5.3.1.2 Post-LOCA Containment Condition (Ref. 10.15) 0-0.5 hr (Cont. Pressure) 63 psia 10.15 0.5-720 hr (Cont. Pressure) 31 psia 5.3.1.3 Isotopic Core Inventory (Ci/MWt) (Ref. 10.45) See Note Below Isotope Ci/MWt Isotope Ci/MWt Isotope Ci/MWt CO-58* 1.529E+02 RU103 4.703E+04 CS136 3.704E+03 CO-60* 1.830E+02 RU105 3.529E+04 CS137 5.626E+03 KR 85 4.711E+02 RUI06 2.259E+04 BA139 4.760E+04 KR 85M 5.908E+03 RHI 105 3.237E+04 BA 140 4.590E+04 RB 86 1.300E+02 SB127 3.379E+03 LA140 4.981E+04 KR 87 1.097E+04 SB129 9.569E+03 LA141 4.325E+04 KR 88 1.539E+04 TE127M 4.508E+02 LA142 4.134E+04 SR 89 2.056E+04 TE127 3.355E+03 CEI41 4.350E+04 SR 90 3.790E+03 TE129M 1.401E+03 CE143 3.910E+04 SR 91 2.677E+04 TE129 9.430E+03 CE144 3.581E+04 SR 92 2.990E+04 TE131M 4.153E+03 PR143 3.783E+04 Y 90 3.981E+03 TE132 3.917E+04 ND147 1.783E+04 Y 91 2.750E+04 1131 2.779E+04 NP239 6.917E+05 Y 92 3.005E+04 1132 3.991E+04 PU238 3.442E+02 Y 93 3.607E+04 1133 5.454E+04 PU239 1.333E+01 ZR 95 4.217E+04 1134 5.937E+04 PU240 2.675E+01 ZR 97 4.419E+04 1135 5.117E+04 PU241 5.419E+03 NB 95 4.237E+04 XE133 5.306E+04 AM241 72.66E+00 MO 99 5.278E+04 XE135 1A82E+04 CM242 2.567E+03 TC 99M 4.621E+04 CS134 1.319E+04 CM244 5.1881E+02

• CO-58 & CO-60 activities are obtained from RADTRAD User's Manual, Table 1.4.3.2-3 (Ref. 10.2)

Note: Additional daughter isotopes added to parent isotopes are shown in Table IA 5.3.1.4 Radionuclide Composition Group Elements Noble Gases Xe, Kr 10.1, RGP 3.4, Table 5 Halogens I, Br Alkali Metals Cs, Rb Tellurium Group Te, Sb, Se, Ba, Sr Noble Metals Ru, Rh, Pd, Mo, Tc, Co Lanthanides La, Zr, Nd, Eu, Nb, Pm, Pr, Sm, Y, Cm, Am Cerium Ce, Pu, Np I Nuclear Common Revision 9 I Nula!omnRvso

CALCULATION CONTINUATION SHEET SHEET 14 of 84 CALC. NO.: H-1-ZZ-MDC-1880

REFERENCE:

LCR H02-01 & DCP 80048085 G. Patel/NUCORE, ORIGINATOR, DATE REV: 11/27/02 1 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 11/29/02 Design Input Parameter Value Assigned Reference 5.3.1.5 Release Fraction (Ref 10.1, Table 1)

BWR Core Inventory Fraction Released Into Containment Gap Release Phase Early In-Vessel Release Phase Group Noble Gases 0.05 0.95 Halogens 0.05 0.25 Alkali Metals 0.05 0.20 Tellurium Metals 0.00 0.05 Ba, Sr 0.00 0.02 Noble Metals 0.00 0.0025 Cerium Group 0.00 0.0005 Lanthanides 0.00 0.0002 5.3.1.6 Timing of Release Phase (Ref. 10.1, Table 4)

Phase Onset Duration Gap Release 2 min 0.5 hr Early In-Vessel Release 0.5 hr 1.5 hr 5.3.1.7 Iodine Chemical Form Iodine Chemical Form  %

Aerosol (CsI) 95.0% 10.1, RGP 3.5 Elemental 4.85%

Organic 0.15%

5.3.1.8 Post-LOCA Drywell Temperature Post-LOCA Time (Hr) Temperature (F) 0 340 Temperature values are bounding 3 320 based on information in Reference 6 250 10.25, pages 35 through 45.

24 208 96 180 240 170 480 150 720 5.3.1.9 Fuel Bumup 58 GWD/MTU <62 GWD/MTU 10.54 & 10.1 5.3.2 Activity Transport in Primary Containment 5.3.2.1 Primary Containment Parameters 5.3.2.2 Drywell Air Volume 169,000 if 710.6.6 & 10.16 Nuclear Common Revision 9

I CALCULATION CONTINUATION SHEET SHEET 15 of 84 CALC. NO.: H-I-ZZ-MDC-1880

REFERENCE:

LCR H02-01 & DCP 80048085 G. Patel/NUCORE, ORIGINATOR, DATE REV: 11/27/02 1 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 11/29/02 Design Input Parameter Value Assigned Reference 5.3.2.3 Suppression Chamber Air 137,000 fiu 10.6.6 & 10.16 Volume 5.3.2.4 Containment Air Volume 306,000 it DI 5.3.2.2 + DI 5.3.2.3 5.3.2.5 Containment Leak Rate 0-24 hrs 0.5 v0/o/day 10.6.4 & 10.15 24-720 hrs 0.25 v0/o/day 10.1, RGP A.3.7 & 10.15 5.3.2.6 Draw Down Time 375 see 10.6.8 5.3.2.7 Cont. Leakage Before Directly Released to 10.1, RGP A.4.2 Draw Down Time (< 375 sec) Environment 5.3.2.8 Cont. Leakage After Directly Released to Reactor 10.1, RGP A.4.2 Draw Down Time (>375 see) Building 5.3.2.9 Reactor Building Volume 4,000,000 ft' 10.6.7 5.3.2.10 Reactor Building Mixing 50% 10.1, RGP A.4.4 5.3.2.11 FRVS Vent Exhaust 9000 din +/- 10% 10.6.3, & 10.20 Rate Before Draw Down 5.3.2.12 FRVS Vent Exhaust 3324 + 5676e6-'8' Actual Eqn in Ref. 10.19, page 24 is Flow Rate After Draw Down 3324 + 5637e6l'1t 5.3.2.13 FRVS Vent Exhaust Filter Efficiency Iodine Species Efficiency (%)

Elemental 900/0 10.47 Aerosol 990/0 Section 7.7 Organic 90% 10.47 5.3.2.14 Post Draw Down FRVS Exhaust Rates For 50% Mixing (using Design Input 5.3.2.12)

Post-LOCA Time (hr) Normal Flow Rate (cfm) 50% Mixing Flow Rate (cfm)

A =3324 + 5676e 18lt A x 1.1 x 2 0 9000 19800 0.104 (375 sec) 9000 19800 0.437 7154 15739 2.104 3860 8492 4.104 3375 7425 8.104 3324 7313 24 3324 7313 96 3324 7313 5.3.2.15 FRVS Recirc Flow Rate 120,000 cfm - 10% 10.6.12 & 10.20 (or, 108,000 efim)

Nuclear Common Revision 9

CALCULATION CONTINUATION SHEET SHEET 16 of 84 CALC. NO.: H-I-ZZ-MDC-1880

REFERENCE:

LCR H02-01 & DCP 80048085 G. Patel/NUCORE, ORIGINATOR, DATE REV- I I27/02 1 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE I1/29/02 Design Input Parameter Value Assigned Reference 5.3.2.16 FRVS Recire Filter Efficiency Iodine Species Efficiency (%)

Elemental 0% Assumed Aerosol 99% Section 7.7 Organic 0% Assumed 5.4 ESF Leakage Model Parameters 5.4.1 Sump Water Volume 118,000 W 10.6.5 & 10.16 5.4.2 ESF Leakage 1 gpm (assumed) 10.18, page 13 5.4.3 ESF Leakage Initiation 0 minute Assumption Time I__

5.4.4 Suppression Pool Water pH >7 10.1, RGP A.2, 10.43, page 11 5.4.5 Sump Water Activity (Ref. 10.1, RGP A.5.1, A.5.3 & Tables 1 & 4)

Group Gap Release Phase Early In-Vessel Release Phase Timing Duration (Hrs) 2 min - 0.50 Hr 0.50 - 2.0 Hr Halogen 0.05 0.25 5.4.6 Iodine Flashing Factor 100/0 10.1, RGP A.5.5 and 10.25, page 35 through 45 5.4.7 Chemical Form Iodine In ESF Leakage Elemental 97% 10.1, RGP A.5.6 Organic 3%

5.5 MSIV Leakage Model Parameters 5.5.1 Total MSIV Leak Rate < 250 scfh 10.6.17 Through All Four Lines 5.5.2 MSIV Leak Rate Through 150 scfh 10.6.17 Line With MSIV Failed 5.5.3 MSIV Leak Rate Through 50 scfh Assumed First Intact Line 5.5.4 MSIV Leak Rate Through 50 scfh Assumed Second Intact Line 5.5.5 Number of Steam Lines 4 10.11 & 10.12e 5.5.6 Diameter and Wall Diameter = 26" 10.13b Thickness of Pipe Between RPV Wall Thickness = 1.117" 10.14c Nozzle & Inboard Isolation Valves HV F022A/B/C/D Nuclear Common Revision 9

CALCULATION CONTINUATION SHEET SHEET 17 of 84 CALC. NO.: H-1-ZZ-MDC-1880

REFERENCE:

LCRH02-01 & DCP 80048085 G. Patel/NUCORE, ORIGINATOR, DATE REV: 1127/02 1 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE I1129/02 Design Input Parameter Value Assigned Reference 5.5.7 Diameter and Wall Diameter = 26" 10.12e Thickness of Pipe Between Wall Thickness = 1.117" 10.14c Inboard & Outboard Isolation Valves HV F028A/B/C/D 5.5.8 Diameter and Wall Diameter = 26" 10.12e Thickness of Pipe Between Wall Thickness = 1.023 10.14a Outboard & 3rd Isolation Valves HV 363 1A/B/C/D 5.5.9 Diameter of Pipe Between Diameter = 28" 10.12a 3rd Isolation & Turbine Stop Wall Thickness = 0.934" 10.14b Valves MSV1/2/3/4 5.5.10 Corrosion Allowance For 0.12" 10.14 Steam 1 5.6 Control Room Model Parameters 5.6.1 CR Volume 85,000ft 10.33, Page 10 5.6.2 CREF System Flow Rate 1,000 cfin 10.6.16 5.6.3 CR Minimum Recirculation 2,600 cfmn 10.6.15 Flow Rate 5.6.4 CR Unfiltered Inleakage 196 +/- 10 cfm actually measured 10.46, page 56 350 cfmn Assumed 5.6.5 CREV System Initiation 30 minutes Assumption 4.8.5 Time After a LOCA 5.6.6 CR Charcoal & HEPA 99% Sections 7.7 & 7.9 Filter Efficiencies 5.6.7 CR Unfiltered Inleakage 500 cfm 10.40, page 6.4-8 & Assumption During Transition 4.8.6 5.6.8 CR Concrete Wall, Floor, and Ceiling Thickness Walls >3 feet 10.27 through 10.31 Floor >3 feet Total Roof Thickness 2'-10-1/2" Ceiling Above CR 1'-0" 10.29a & 10.29b 5.6.9 CR X/Qs For Containment & ESF Leakage Release Via FRVS Vent Ground Level Release Time XIQ (see/rm) 0-2 1.25E-03 10.5, page 34 2-8 8.09E-04 8-24 3.04E-04 24-96 2.10E-04 96-720 1.59E-04 Nuclear Common Revision 9I

Design Input Parameter I Value Assigned Reference 5.6.10 CR X/Qs For MSIV Leakage Release Via Turbine Building Louvers Ground Level Release Time X/Q (sec/mr3 )

0-2 6.17E-04 10.5, page 35 2-8 4.OOE-04 8-24 1.44E-04 24-96 1.OOE-04 96-720 7.49E-05 5.6.11 CR Occupancy Factors Time (Hr)  %

0-24 100 10.1, RGP 4.2.6 24-96 60 96-720 40 5.6.12 CR Breathing Rate 3.5E-04 (ms/sec) 10.1, RGP 4.2.6 5.6.13 Minimum Reactor Bldg 1'-6" 10.35 Wall Thickness T 5.7 Site Boundary Release Model Parameters 5.7.1 EAB X/Q (0-2 Hrs) 1.9E-04 see/ms 10.32, pages 5 & 9 5.7.2 EAB Breathing Rate ] 3.5E-04 mP/sec 110.1 5.7.3 LPZ X/Qs (0-720 Hrs)

Time XIQ (sec/im) 0-2 1.9E,-05 10.32, pages 5 & 9 2-4 1.2E-05 4-8 8.OE-06 8-24 4.OE-06 24-96 1.7E-06 96-720 4.7E-07 5.7.4 Offsite Breathing Rates Time BR (ml/sec) 0-8 3.5E-04 10.1, RGPs 4.1.3 & 4.4 8-24 1.8E-04 24-720 2.3E-04 5.7.5 CR Charcoal Filter Dimensions Approximated Conservatively Length 3 feet 10.38 Height 3 feet Width 4 feet Nuclear Common Revision 9

CALCULATION CONTINUATION SHEET SHEET 19 of 84 CALC. NO.: H-I-ZZ-MDC-1880

REFERENCE:

LCR H02-01 & DCP 80048085 G. Patel/NUCORE, ORIGINATOR, DATE REV: 11/27/02 1 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 11/29/02 Design Input Parameter Value Assigned Reference 5.7.6 Charcoal Density 0.70 g/cc Assumed 5.7.7 Concrete Density 2.3 g/cc Assumed 5.7.8 Dose Point Location 143'-0" 6' above EL 137'-0"

__________ I__________ I__________

Revision 9 I Nuclear Common I Nuclear Common Revision 9 1

CALCULATION CONTINUATION SHEET SHEET 20 of 84 CALC. NO.: H-I-ZZ-MDC-1880

REFERENCE:

LCR H02-01 & DCP 80048085 G. Patel/NUCORE, ORIGINATOR, DATE REV: 11/27/02 1 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 11/29/02 6.0 METHODOLOGY The design basis loss of coolant accident is analyzed using a conservative set of assumptions and as-built design inputs to demonstrate the performance of one or more aspects of the facility design to protect the control room operator and the health and safety of the general public. The guidance in Regulatory Guide 1.183 (Ref. 10.1) is followed along with the plant-specific design input parameters computable for the AST and TEDE dose criteria. The numeric values of the post-accident performance of ESF components are conservatively selected to assure an appropriate and prudent safety margin against unpredicted events in the course of an accident and compensate for large uncertainties in facility parameters, accident progression, radioactive material transport, and atmospheric dispersion.

6.1 Post-LOCA Containment Leakage:

6.1.1 Source Term:

The post-LOCA containment leakage model is shown in Figure 1. The core inventory listed in Table 1 is released into the containment at the release timing and fractions shown in Tables 3 & 4 (Ref. 10.1, RGPs 3.2 & 3.3). Since the post-LOCA minimum suppression chamber water pH is greater than 7.0 (Ref.

10.43), the chemical form of radioiodine released into the containment is assumed to be 95% cesium iodide (CsI), 4.85 percent elemental iodine, and 0.15 percent organic iodide as shown in Table 5. With the exception of elemental and organic iodine and noble gases, the remaining fission products are assumed to be in particulate form (Ref 10.1, RGP 3.5). The plant-specific uprated core is developed in Table IA from Reference 10.54. As recommended in RADTRAD Table 1.4.3.3-2, the inventories listed for some of the parent isotopes include their significant daughter products. The final composite core inventory is shown in Table lB. The RADTRAD Nuclide Inventory File (NIF) HEPUMHA_def.txt is developed based on the uprated core inventory in Table lB and used for the containment, ESF, and MSIV leakage paths. The source term design inputs are shown in Sections 5.3.1.1 through 5.3.1.8.

6.1.2 Transport In Primary Containment:

The radioactivity released from the fuel is assumed to mix instantaneously and homogeneously throughout the free air volume of the primary containment as it is released. The radioactivity release into the containment is assumed to terminate at the end of the early in-vessel phase, which occurs at the end of 2 hrs after the onset of a LOCA (Ref.10.1, Table 4). The design inputs for the transport in the primary containment are shown in Sections 5.3.2.1 through 5.3.2.9.

6.1.3 Reduction In Airborne Activity Inside Containment The airborne iodine and aerosol are removed from the reactor building environment by the FRVS recirculation system, which re-circulates air at a design rate of 108,000 cfm or 1.62 vol/hr (108,000 ft3 /min x 60 min/hr x (4.OOE+06 ft)" 1 = 1.62 vol/br). Although, the FRVS recirculation system provides a good mixing of activity in the reactor building (RB), the airborne activity is conservatively assumed to mix with only 50% of the RB volume (Ref. 10.1, RGP A.4.4.). To simulate the 50% mixing in the RB, I Nuclear Common Revision 9 1 I Nula omnRvso

CALCULATION CONTINUATION SHEET SHEET 21 of 84 CALC. NO.: H-I-ZZ-MDC-1880

REFERENCE:

LCR H02-01 & DCP 80048085 G. Patel/UCORE, ORIGINATOR, DATE REV: 11/27/02 1 M. Drucker/NUCORE, REVIEWERNVERIFIER, DATE 11/29/02 the exhaust rate of the FRVS vent system is doubled as shown in Design Input 5.3.2.14. The FRVS vent exhaust rate varies with time as shown by the equation in Design Input 5.3.2.12. Design Input 5.3.2.14 provides the FRVS exhaust flow rates at 100% and 50% mixings. The airborne activity in the RB is reduced by both the FRVS recirculation and FRVS vent filtration system before it is released to the environment. The charcoal and HEPA filtration efficiencies are shown in Section 5.3.2.16.

6.1.4 Dual Containment:

Leakage from the primary containment is assumed to be released directly to the environment prior to draw down time during which the RB does not maintain a negative pressure as defined in technical specifications (Ref 10.1, RGP A.4.2). 50% mixing is credited for dilution of the activity in the RB (Ref.

10.1, RGP A.4.4). The containment leakage RADTRAD input and output files are listed in the Attachments B and C and the EAB, LPZ, and CR TEDE doses are shown in the Section 8.0.

6.1.5 Containment Purging:

The HCGS containment is not purged for combustible gas or pressure control measure within 30 days of the LOCA. Therefore, the release containment purging is not analyzed per RG 1.183, RGP A.7.

6.2 Post-LOCA ESF Leakage:

The post-LOCA ESF leakage release model is shown in Figure 2. The ESF systems that recirculate suppression pool water outside of the primary containment are assumed to leak during their intended operation. This release source includes leakage through valve packing glands; pump shaft seals, flanged connections, and other similar components. The radiological consequences from the postulated leakage is analyzed and combined with consequences from other fission product release paths to determine the total calculated radiological consequences from the LOCA (see Section 8.0 of this calc). The ESF components are located in the RB.

6.2.1 Source Term:

With the exception of noble gases, all the fission products released from the fuel to the containment (as defined in Sections 5.3.1.3 & 5.3.1.5) are assumed to instantaneously and homogeneously mix in the suppression pool water at the time of release from the core. The total ESF leakage from all components in the ESF recirculation systems is 1 gpm. This ESF leakage is doubled (Ref 10.1, RGP A.5.2) and assumed to start at time t=0.0 minute after onset of a LOCA. With the exception of iodine, all remaining fission products in the recirculating liquid are assumed to be retained in the liquid phase. The design inputs for the ESF leakage are shown in Section 5.4.

I Nuclear Common Revision 9 1 I Nula omnRvso

CALCULATION CONTINUATION SHEET SHEET 22 of 84 CALC. NO.: H-I-ZZ-MDC-1880

REFERENCE:

LCR H02-01 & DCP 80048085 G. Patel/NUCORE, ORIGINATOR, DATE REV: I1/27102 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 11/29/02 6.2.3 Chemical Form The radioiodine that is postulated to be available for release to the environment is assumed to be 97%

elemental and 3% organic (Ref. 10.1, RGP A.5.6). The reduction in ESF leakage activity by dilution in the RB and removal by FRVS recirculation and FRVS vent filtration systems are credited.

The ESF leakage RADTRAD inputs and outputs files are listed in the Attachments D & F and the EAB, LPZ, and CR TEDE doses are shown in the Section 8.0.

6.3 Post-LOCA MSIV Leakage:

The main steam isolation valves (MSIVs) have design leakage that may result in a radioactivity release.

The radiological consequences from postulated MSIV leakage are analyzed and combined with consequences postulated for other fission product release paths to determine the total calculated radiological consequences from the LOCA. The following assumptions are acceptable for evaluating the consequences of MSIV leakage.

6.3.1 Source Term For the purpose of this analysis, the activity available for release via MSIV leakage is assumed to be that activity released in the drywell for evaluating containment leakage.

A total of 250 scfh (the maximum proposed allowable leakage limit) is assumed to occur as follows based on the NRC accepted method of splitting the total MSIV leakage (Ref. 10.51, page 6):

(1) 150 scfh through the steam line with the failed MSIV. The plate out of activity and holdup time are not credited in the steam line between the inboard and outboard valves. The plateout and holdup are credited in the steam lines from the RPV to inboard isolation valve, outboard isolation valve to turbine block valve, which is conservative.

(2) 50 scfh through a first intact steam line. The plate out of activity and holdup time are credits in the entire steam line from the RPV nozzle to turbine stop valve.

(3) 50 scfh through a second intact steam line. The plate out of activity and holdup time are credits in the entire steam line from the RPV nozzle to turbine stop valve.

The staff allows arbitrarily distributing the remainder MSIV leakage that makes up the difference between the maximum allowable leakage per line and the total leakage based on the following inherent conservatisms that exist in the MSIV leakage path.

I. As-Tested Configuration of MSIV Leakage Paths The MSIV Type C leak rate test is performed to comply with the T.S.4.6.1.2.f and T.S.3.6.1.2.c by pressurizing the pipe spool between the inboard and outboard MSIVs and measuring the total leakage I Nuclear Common Revision 9 1I I Nula omnRvso

CALCULATION CONTINUATION SHEET SHEET 23 of 84 CALC. NO.: H-I-ZZ-MDC-1880

REFERENCE:

LCR 1102-01 & DCP 80048085 G. Patel/NUCORE, ORIGINATOR, DATE REV: 11/27/02 1 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 11/29/02 through both MSIVs (Refs 10.52 & 10.53). The analyzed worst case is when maximum allowed leakage of 150 scfh occurs in one main steam line with the inboard MSIV failed. The leakage out that line is assumed to be 150 scfh.

The Tech Spec allowed remaining MSIV leakage is 100 scfh (250 scfh - 150 scfh = 100 scth). If 100 scfh occurs in one of the remaining 3 main steam lines to maximize the environmental release dose, the analyzed maximum leakage cannot be 100 scfh. The Tech Spec testing measures the total leakage between the MSIVs. The LLRT does not measure the leakage through individual MSIV but it only determines the total leakage (Ref. 10.53). For this 100 scfh line, if one MSIV (inboard or outboard) leaks at 100 scth then the other MSIV has zero leakage and the line leakage is zero (the dose calc case is non-conservative). The worst case is with both the inboard and outboard MSIVs leaking at 50 scfh for a line leakage of 50 scfh (even though the test boundary leakage would be 100 scfh).

The MSIV leakages occur through four main steam lines and it is considered a single leakage path for analysis. This is based on a single assumed failure of the inboard MSIV with the outboard MSIV having the maximum leakage (150 scfh). The MSIV leakage of 150 scfh in the failed MSIV line and 50 scfh in two intact lines is conservative based on the as-tested configuration of the MSIV leakage path. The MSIV leakage of 50 scfh in the second and third intact lines is conservative. Therefore, the current analysis meets the as-tested configuration of the MSIV leakage paths for total leakage of 250 scfh and a maximum of 150 scfh for a single line. Using leakages of 150 (failed MSIV line), 50, 50 and 0 as the leakage through the 4 steam lines is conservative. Based on the testing method using 150 scfh (failed MSIV line), 100, 0, and 0 scfh is not necessary for this analysis.

2. Assumption of Constant MSIV Leak Rate The Hope Creek MSIV leakage path involves three MSIVs in the series. The maximum MSIV leakage of 150 scfh occurs at the post-LOCA drywell design pressure of 63 psia (Ref. 10.15) at the outboard MSIV assuming the inboard MSIV failed to close to comply with the single failure criterion. The actual MSIV leakage (acth) will be approximately 1/4 of scfh because the post-LOCA drywell pressure is 4 times the atmospheric pressure. The MSIV leakage to environment from the failed MSIV line is further limited by the main steam stop valve (MSSV), which in series with the outboard MSIV. If the MSSV is leaking at a lower rate, the MSIV leakage from the failed MSIV line will be further reduced accordingly.

To postulate the MSIV leakage of 150 seth from the failed MSIV line, the MSSV should leak at a rate of 150 scfh, which is highly improbable. However, the MSSVs are currently not in the Type C Test Program since the MSIVSS is deleted, they were tested for low leakages. The same arguments can be applied to reduce the MSIV leakage through the remaining intact main steam lines.

The MSIV leakage is assumed to continue for entire duration of the accident. Per RG 1.183, RGP A.6.2 (Ref. 10.1), the MSIV leakage is reduced to a value equal to 50% of the maximum leak rate after the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, based on the post-LOCA drywell pressure (Ref. 10.15).

The aerosols in the MSIV leakage settle down in the main steam line due to gravitational deposition.

The aerosol removal from the MSIV leakage is calculated in Section 7.4 using the NRC approved method in Reference 10.22, which uses the Monte Carlo distribution of aerosol settling velocity in well I Nuclear Common Revision 9 i

CALCULATION CONTINUATION SHEET SHEET 24 of 84 CALC. NO.: H-I-ZZ-MDC-1880

REFERENCE:

LCR H02-01 & DCP 80048085 G. Patel/NUCORE, ORIGINATOR, DATE REV: I1/27/02 1 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE I It29/02 mixed flow. The analysis in Section 7.4.1 takes the credit of volumetric flow rates and volumes of main steam piping upstream and down stream of inboard isolation valves because the steam lines from the RPV nozzle to turbine stop valve are seismically designed and supported for Safe Shutdown Earthquake (SSE) (Ref 10.26 & 10.37). The analysis in Section 7.4.1 determines that a large amount of airborne aerosol in MSIV leakage will be deposited on the steam pipe surface.

The reduction in elemental iodine activity in the MSIV leakage is calculated in Section 7.4.2 using the staff recommended guidance on acceptable method in Reference 10.23 (Reference A-9 of RG 1.183).

Both, the temperature dependent elemental iodine deposition and resuspension rates, net iodine deposition rates, and iodine removal efficiencies are calculated in Tables 2 through 9 using J.E. Cline method (Ref 10.23). The remaining airborne activity in the MSIV leakage after removal by deposition (aerosol) and plateout (elemental) mechanisms is directly released to the environment as a ground level release through the turbine building louvers.

The holdup times for each MSIV leakage release path (MSIV failed and intact steam lines) are calculated in Sections 7.2 and 7.3 based on the leakage rates and well-mixed steam piping volumes.

These parameters calculated in Sections 7.2, 7.3, & 7.4 are input in the RADTRAD MSIV release model to calculate EAB, LPZ, and CR doses, which are listed in Section 8.0. The design inputs for the MSIV leakage are shown in Section 5.5.

6.4 Control Room Model The post-LOCA control room RADTRAD nodalization is shown in Figure 4 with the design input parameters. The post-LOCA radioactive releases that contribute the CR TEDE dose are as follows:

• Post-LOCA Containment Leakage

• Post-LOCA ESF Leakage

• Post-LOCA MSIV Leakage The radioactivity from the above sources are assumed to be released into the atmosphere and transported to the CR air intake, where it may leak into the CR envelope or be filtered by the CR intake and recirculation filtration system and distributed in the CR envelope. There are four major radioactive sources, which contribute to the CR TEDE dose are:

• Post-LOCA airborne activity inside the CR 0 Post-LOCA airborne cloud external to CR 0 Post-LOCA containment shine to CR

• Post-LOCA CREF filter shine 6.4.1 Post-LOCA Airborne Activity Inside CR The post-LOCA radioactive releases from various sources are discussed in Sections 6.1 through 6.3 above and shown in Figure 4. The activities releases from the various sources are diluted by the I Nuclear Common Revision 9 1

CALCULATION CONTINUATION SHEET SHEET 25 of 84 CALC. NO.: H-1-ZZ-MDC-1880

REFERENCE:

LCR H02-01 & DCP 80048085 G. Patel/NUCORE, ORIGINATOR, DATE REV: 11/27/02 1 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 11/29/02 atmospheric dispersions and carried to the CR air intake. The atmospheric dispersion factors are shown in Sections 5.6.9 & 5.6.10 for the containment/ESF and MSIV leakages. The containment and ESF leakages have the same release point (FRVS vent) and X/Qs. The RADTRAD release models are developed for each release path using appropriate design inputs from Sections 5.3, 5.4, and 5.5. The CR dose model is developed using the design input parameters in Section 5.6. The CR airborne TEDE dose contributions from the above post-LOCA sources are calculated and tabulated in Section 8.0.

6.4.2 Post-LOCA Airborne Cloud External to CR The radioactive plumes released from various post-LOCA sources are carried over the CR building, submerging the CR in the radioactive cloud. The CR operator is exposed to direct radiation from the radioactive cloud external to the CR structure. The review of control building concrete structure drawings (Ref. 10.27 through 10.31) indicate that the CR is surrounded by at least 2'-10-1/2" (1' ceiling

@ EL 155'-3" and 1'-10-1/2" roof@ EL 172'0") concrete shielding with a minimum distance of 29 feet from the least shielding (172'-0" - (137'-0" + 6'-0" tall person)). This minimum-shielding configuration provides an adequate protection to the CR operator to reduce the CR operator external cloud dose to a negligible amount.

6.4.3 Post-LOCA Containment Shine to CR The post-LOCA airborne activity in the containment is released into the reactor building (RB) via containment leakage through the penetrations and openings and gets uniformly distributed inside the RB. The airborne activity confined in the dome space of the RB contributes direct shine dose to the CR operator. The review of the containment building concrete structure drawing (Ref. 10.35) indicates that the minimum dome concrete thickness is 1'-6". The CR minimum roof/ceiling concrete shielding is 2'-

10-1/2". The combined concrete shielding of 4'-4-1/2" (1'-6" + 2'-10-1/2" = 4'-4-1/2") provides ample shielding to reduce the CR operator containment shine dose to an insignificant amount.

6.4.4 Post-LOCA CREF Filter Shine The two trains of CREF charcoal and HEPA filters are located above the CR operating floor at elevation 155'-3" (Refs. 10.28, 10.29, & 10.39). The CR operating floor is located at elevation 137'-0" (Ref.

10.28c). The concrete floor at EL 155'-3" is 1 feet thick (Ref. 10.29). The filter assembly is placed on a 6" concrete pad (Ref. 10.39c, Section DD), which provides the total concrete shielding of 1'-6" between the CR operator and charcoal/HEPA filter. The receptor location is assumed to be located at 6 feet above the CR operating floor right below the center of a charcoal filter. The iodine and aerosol activities are conservatively collected on the charcoal bed. The dimensions of charcoal filter housing are obtained from Reference 10.38 and are conservatively approximated to 3' (L) x 3' (HI)x 4' (W) by summing all of the charcoal filter trays within a filter housing as shown in Figure 5, which also shows the dose point location.

I Nuclear Common Revision 9 1I I ula omnRvso

I CALCULATION CONTINUATION SHEET SHEET 26 of 84 CALC. NO.: H-1-ZZ-MDC-1880

REFERENCE:

LCR H02-01 & DCP 80048085 G. Patel/UCORE, ORIGINATOR, DATE REV: 11/27/02 1 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 11/29/02 Post-LOCA Activity The RADTRAD3.02 code does not provide the post-LOCA iodine and aerosol activities accumulated on charcoal and HEPA filters. Therefore the iodine and aerosol activities are conservatively calculated as follows:

1. The time dependent isotopic iodine and aerosol integrated activities in the CR due to the post-LOCA containment leakage are calculated without taking credit for the CR charcoal and HEPA filters (RADTRAD File HAST1000CL03.psf). The time dependent integrated activities are shown as Case 1 in Table 11.

2. The time dependent isotopic iodine and aerosol integrated activities in the CR due to the post-LOCA containment leakage are calculated with taking credit for the CR charcoal and HEPA filters (RADTRAD File HAST1000CL02.psf). The time dependent integrated activities are shown as Case 2 in Table 10.

3. The total isotopic iodine and aerosol activities on the CR filters due to the containment leakage are calculated in Table 12 (i.e., Case 1 - Case 2).

4. Similarly, the time dependent isotopic iodine and aerosol integrated activities in the CR due to the post-LOCA MSIV and ESF leakages are calculated in Tables 13 - 14 and in Tables 15 - 16, respectively.

5. The total integrated iodine and aerosol activities on the CR filters shown in Table 17 were calculated in Revision 0 of this LOCA dose calculation based on a CR unfiltered inleakage of 1,000 cfm and an ESF leakage of 10 gpm. The CR filter shine dose calculated in Revision 0 of this LOCA dose calculation is negligible (see Section 8.0 of this calculation). In this revision of the LOCA dose calculation, the CR unfiltered inleakage and ESF leakage are reduced to 350 cfm and 1 gpm respectively. Table 17 indicates that the accumulation of iodine on charcoal bed contributes the major shine dose. The activity accumulated on the charcoal filter and the resulting charcoal filter shine dose are functions of CR unfiltered inleakage rate and source strength of airborne activity. The reductions in the CR unfiltered inleakage and ESF leakage will reduce the CR charcoal filter shine dose substantially (see Table 17 for activities from the different sources), which will compensate any increase in the airborne iodine activity due to deletion of the FRVS charcoal recirculation filtration, reduction in the FRVS vent charcoal filter efficiency, and increase in the core uprated core inventory. Therefore, the previously calculated CR filter shine dose is judged to be bounding for the subject changes.

6. The total isotopic activities on the CR charcoal filter bed in Table 17 are input into the MicroShield Computer code (Ref. 10.9) with the source geometry, dimension, and detector location to compute the direct dose rate from the CR filter. The direct dose from the CR filter shine is calculated in Section 7.6 using the CR occupancy factors.

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CALCULATION CONTINUATION SHEET SHEET 27 of 84 CALC. NO.: H-I-ZZ-MDC-1880

REFERENCE:

LCR H02-01 & DCP 80048085 G. Patel/NUCORE, ORIGINATOR, DATE REV: 11/27/02 1 M. DruckeriNUCORE, REVIEWERNERIFIER, DATE 11/29/02 6.5 CR & FRVS Vent Charcoal/IEPA Filter Efficiencies The CR and FRVS vent charcoal filters are tested to comply with Generic Letter 99-02 requirements (Refs. 10.3 & 10.6). However, there is no specific criteria to establish the HEPA filter efficiency, the GL 99-02 criterion is used to determine the HEPA filter efficiency. The in-place penetration testing acceptance requirements are given in Hope Creek Technical Specifications (Ref. 10.6). The filter efficiencies credited in this analysis are calculated in Section 7.7 based on the testing criteria in Reference 10.6 and GL 99-02 (Ref. 10.3).

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CALCULATION CONTINUATION SHEET SHEET 28 of 84 CALC. NO.: H-I-ZZ-MDC-1880

REFERENCE:

LCR H02-01 & DCP 80048085 G. Patel/NUCORE, ORIGINATOR, DATE REV: 11/27/02 1 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 11/29/02 7.0 CALCULATIONS 7.1 HCGS Plant Specific Nuclide Inventory File (NIF) For RADTRAD3.02 Input The RADTRAD nuclide inventory file Bwr.def_NIF establishes the power dependent radionuclide activity in Ci/MWt for the reactor core source term. Since these core radionuclide activities are dependent on the core thermal power level, reload design, and burnup, the NIF is modified based on the plant-specific uprated core information obtained from Reference 10.45. The RADTRAD NIF HEPUMHADEF.txt is modified based on the uprated core inventory in Table IB and used in the analyses.

7.2 Main Steam Line Volumes & Surface Area For Plateout of Activity 7.2.1 MSIV Line Between RPV Nozzle & Inboard Isolation Valve Piping Class = DLA (Ref. 10.13b)

Pipe Diameter = 26" (Ref. 10.13b)

Minimum Wall Thickness = 1.117" (Ref. 10. 14c)

Corrosion Allowance For Steam = 0.12" (Ref. 10.14c)

Total Minimum Thickness = 1.117" + 0.12" = 1.237" 26" Pipe ID = OD - (2 x Min Wall Thickness) = 26" -2 x 1.237" = 23.526" = 1.961' Shortest Length of Pipe Between RPV Nozzle & Inboard Isolation Valves a 91 ' (Ref. 10.13)

Volume of Pipe Between RPV Nozzle & Inboard Isolation Valves

=A 2Ar L= 7 (1.961/2) 2 x 91' = 1 274.84 ft3 =7.79 m3 I Pipe Surface Area = n D L-=,x 1.961' x 91' I560.62 t 52.11 m 2 7.2.2 MSIV Line Between Inboard & Outboard Isolation Valves Piping Class = DLA (Ref. 10.13b)

Pipe Diameter = 26" (Ref. 10.1 3b)

Minimum Wall Thickness = 1.117" (Ref. 10.14c)

Corrosion Allowance For Steam = 0.12" (Ref. 10. 14c)

Total Minimum Thickness = 1.117" + 0.12" = 1.237" I Nuclear Common Revision 99 1I Revision I Nuclear Common

CALCULATION CONTINUATION SHEET SHEET 29 of 84 CALC. NO.: H-I-ZZ-MDC-1880

REFERENCE:

LCR H02-01 & DCP 80048085 G. Patel/NUCORE, ORIGINATOR, DATE REV: 11/27/02 1 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 11/29/02 26" Pipe ID = OD - (2 x Min Wall Thickness) = 26" -2 x 1.237" = 23.526" = 1.961' Length of Pipe Between Inboard & Outboard Isolation Valves = 25.67' (Ref. 10.13)

Volume of Pipe Between Inboard & Outboard Isolation Valves

= r3LA = n(1.961/2)2 x 25.67 = 1 77.53 f3 = 2.2 m'3 Pipe SurfaceArea=xD L==xx 1.961' x 25.67' = 158.14 ft2 = 14.70 m2 7.2.3 MSIV Line Between Outboard & Third Isolation Valves Piping Class = DBB (Ref. 10.11 & 10.12e)

Pipe Diameter = 26" (Ref. 10.12e)

Minimum Wall Thickness = 1.023" (Ref. 10.14a)

Corrosion Allowance For Steam = 0.12" (Ref. 10.14a)

Total Minimum Thickness = 1.023" + 0.12" = 1.143" 26" Pipe ID = OD - (2 x Min Wall Thickness) = 26" - 2 x 1.143" = 23.714" = 1.976' Length of Pipe Between Outboard & Third Isolation Valves = 41.33' (Ref. 10.12e)

Volume of Pipe Between Outboard & Third Isolation Valves

= 7C r2 L = 7t (1.976/2) 2 x 41.33' = 1 126.74 t3 = 3.59 m3 Pipe Surface Area = 7 D L =,x x 1.976' x 41.33' = 1 256.57 ft2 23.85 m2 7.2.4 MSIV Line Between Third Isolation Valve and Turbine Stop Valve Piping Class = DBC (Ref. 10.11 & 10.12a)

Pipe Diameter= 28" (Ref. 10.12a)

Minimum Wall Thickness = 0.934" (Ref. 10.14b)

Corrosion Allowance For Steam = 0.12" (Ref. 10.14b)

Total Minimum Thickness =0.934" + 0.12" = 1.054" 28" Pipe ID = OD - (2 x Min Wall Thickness) = 28" - 2 x 1.054" = 25.892" = 2.158' Length of Pipe Between Third Isolation Valve & Turbine Stop Valve = 272.6' (Ref. 10.12a)

I Nuclear Common Revision 9 1I I ula omnRvso

Volume of Pipe Between Third Isolation Valve & Turbine Stop Valve

= 7 r2L (2.158/2)2 x 272.6' = 1997.05 ft = 28.26 m I 2

1848.11ft= 171.78m Pipe Surface Area = 7 D L = 7r x 2.158' x 272.6' =

7.2.5 Surface Area & Volume of Failed MSIV Steam Line Total Volume of MSIV Leakage Path For MSIV Failed Steam Line is conservatively calculated by neglecting the volume of steam line between the inboard and outboard isolation MSIVs.

= 274.84 ft3 + 126.74 ft + 997.05 ft3

- 1,398.63 f 3 = 39.58 m3 1 Total Surface Area of MSIV Leakage Path For MSIV Failed Steam Line

= 560.62 f 2 + 256.57 ft2 + 1,848.11 ft2

= 2,665.56 ft2 = 247.77 m2 7.2.6 Surface Area & Volume of Intact Steam Lines Total Volume of MSIV Leakage Path For Intact Steam Lines 3 3

- 274.84 ft 3 + 77.53 W + 126.74 fW + 997.05 ft 3

1,476.16 t = 41.83 m Total Surface Area of MSIV Leakage Path For Intact Steam Lines

= 560.62 ft 2 + 158.14 ft2 +256.57 ft2 + 1,848.11 ft2

= 2,823.7 ft2 262.46 in2 7.3 Holdup Times MSIV Leak Rate of 250 sefh Holdup Time for MSIV Leakage of 150 scfh for each of two MSIV Failed Line

= 1,398.63 ft3 / 150 ft3/hr = 9.32 hrs I Nuclear Common Revision 9 I I ula omnRvso

CALCULATION CONTINUATION SHEET SHEET 31 of 84 CALC. NO.: H-1-ZZ-MDC-1880

REFERENCE:

LCR H02-01 & DCP 80048085 G. Patel/NUCORE, ORIGINATOR, DATE REV: 11/27/02 1 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 11/29/02 Holdup Time for MSIV Leakage of 50 scfli/MSIV For MSIV Intact Lines

= 1,476.16 f& / 50 f 3 /hr = 29.52 hrs MSIV Leak Rate For MSIV Failed Line = 150 &/hr x 1/60 hr/min = 2.50 cfin 3 /hr x 1/60 hr/min = 0.8334 cfm MSIV Leak Rate For MSIV Intact Lines = 50 fO MSIV leakage rate is halved after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 7.4 Plateout of Activity in Main Steam Lines 7.4.1 Aerosol Deposition Reference 10.37 indicates that the HCGS main steam piping from the reactor pressure vessel (RPV) nozzle to the turbine stop valve is seismically analyzed to assure the piping wall integrity during and after a seismic (safe shutdown earthquake [SSE]) event. The Hope Creek turbine building is classified as Non-seismic, however, codes and criteria similar to those for Seismic Category I structure, were used for the structure design of the entire building (Ref. 10.26, Section 1.2). The turbine building was dynamically analyzed and design to accommodate an SSE event (Ref. 10.37, page 1-2) so that it does not collapse on, or interact with, adjacent seismic Cat I structures for SSE. 10 CFR Part 100 requires that the structures, systems, and components necessary to ensure the capability of mitigating the radiological consequences of an accident that could result in exposures comparable to the does guideline of Part 100 be designed to remain functional during and following a safe-shutdown earthquake. The main steam lines and housing structures are qualified to meet Part 100 requirements; therefore, the main steam lines are credited for the aerosol deposition and holdup for MSIV leakage path.

The Brockmann model for aerosol deposition (Ref. 10.2, Section 2.2.6.1) is based on the plug flow model. The staff concluded that the plug flow model for aerosol deposition in the main steam piping under-predicts the dose (Ref 10.22, Appendix A). The aerosol settling velocity in the well-mixed flow model depends on the variables having a large range of uncertainty (see Equation 5 of Appendix A of Ref. 10.22). Therefore, the following aerosol deposition model is used, which is accepted by the Staff in Reference 10.22, Appendix A). Therefore, the Staff performed a Monte Carlo analysis to determine the distribution of aerosol settling velocities for the main steam line during the in-vessel release phase.

The accepted 40 percentile settling velocity is reasonably conservative for aerosol deposition in the MSIV leakage. The results of the Monte Carlo analysis for settling velocity in the main steam line are given in the following Table:

Percentile Settling Velocity Removal Rate (m/see) Constant (hr"')

60'_ (average) 0.00148 11.43 50 (median) 0.00117 9.04 40r 0.00081 6.26 10l 0.00021 1.62 I Nuclear Common Revision 9 1

CALCULATION CONTINUATION SHEET SHEET 32 of 84 CALC. NO.: H-I-ZZ-MDC-1880

REFERENCE:

LCR H02-01 & DCP 80048085 G. Patel/NUCORE, ORIGINATOR, DATE REV: I It27/02 1 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE I It29/02 7.4.1.1 MSIV Failed Line The derivation of staff's well-mixed model begins with a mass balance as follows (Ref.10.22, Page A-2):

V* dC=Q

• Ci-Q* C-*v*C (1) dt Where V = volume of well-mixed region C = concentration of nuclides in volume Q = volumetric flow rate into volume A, = rate constant for settling And X fi_

V Where us = settling velocity A = settling area Under steady-state condition, the derivation in the above equation (1) becomes zero. Equation (1) can be simplified as follows:

C ---Cin I 1+ X1*V Q

RADTRAD allows input of filter efficiency for each flow path. Noting that C is also the concentration of nuclide leaving the volume, the above equation can be used to determine equivalent filter efficiency as follows:

1lf'jt= I -C =1 - 1 (2)

Cin 1+ X.**V Q

Using equation (2), the aerosol removal efficiencies can be calculated for the MSIV failed and intact steam lines as follows:

For MSIV Failed Line 40 percentile settling velocity of 8.1E-04 results in the aerosol removal rate of 6.26 hri1 (Ref. 10.22, Page A-3)

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I CALCULATION CONTINUATION SHEET ISHEET 33 of 84 CALC. NO.: H-I-ZZ-MDC-1880

REFERENCE:

LCR H02-01 & DCP 80048085 G. Patel/NUCORE, ORIGINATOR, DATE REV: 11/27/02 1 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 11/29/02

= 6.26 hr ' V = 1,398.63 ft Q =150 f?/r Substituting values in equation (2) yields:

11fiR=1- 1 1+ &*V Q

flflt = 1- 1 I ....

1 + 6.26 1/hr

• 1.398.63 if 150 ft3/hr Tlnitlt1- 1 - 1 = 1 - 0.0168 = 0.9832 or 98.32%

1 + 58.37 59.37 7.4.1.2 MSIV Intact Lines 3 Q =50 ft/hr

ý,= 6.26 hf' V = 1,476.16 Wt Substituting values in equation (2) yields:

T1flit 1- 1 1+ S*V Q

Tlfilt = 1 - 1 1+ 6.261/hr* 1,476.16Yf1 50 fe/hr 11fit = 1 - 1 = 1- 1 = 1 - 0.00538 = 0.9946 or 99.46%

1 + 184.82 185.82 7.4.2 Elemental Iodine Gaseous iodine tends to deposit on the piping surface by chemical adsorption. The elemental iodine being the most reactive has the highest deposition rate. The iodine deposited on the surface undergoes both physical and chemical changes and can be re-emitted as an airborne gas (re-suspension) or permanently fixed to the surface (fixation). The RGP A.6.5 (Ref. 10.1) indicates that Reference A-9 provides acceptable models for deposition of iodine on the pipe surface. Reference 10.23, which is Reference A-9 of Regulatory Guide 1.183 is used to determine the deposition and resuspension rates of elemental iodine as follows:

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CALCULATION CONTINUATION SHEET SHEET 34 of 84 CALC. NO.: H-I-ZZ-MDC-1880

REFERENCE:

LCR H02-01 & DCP 80048085 G. Patel/NUCORE, ORIGINATOR, DATE REV: 11/27/02 1 M. Drucker/NUCORE, REVIEWERVERIFIER, DATE 11/29/02 di = elemental iodine vapor deposition velocity (cm/s) = e(2 809 Fr - 12.80 (+ 0.33)) = e(2809/T - 12-5) (Ref 10.23, pages 4 & 12).

Where T = gas temperature (OK)

This equation is same as equation 30 in Bixler Model in the RADTRAD3.02 code (Ref. 10.2, page 212).

The elemental iodine deposition velocities are calculated in Table 2 based the post-LOCA drywell temperature shown in Design Input 5.3.1.8.

The elemental iodine deposition rate 4~d (Vr") = di

• S *3600 (Ref. 10.23, page 4)

V Where di = deposition velocity (mn/sec)

S = surface area of deposition (m2 )

V = volume (m3)

The deposition velocity in cm/sec (which is converted into ra/sec) and elemental iodine deposition rates at various drywell temperatures are calculated in Tables 3 & 4 for the MSIV failed and intact steam lines respectively.

The portion of elemental iodine deposited on the pipe surface will be resuspended as an airborne gas (organic iodine). Since the CR filtration efficiencies are same for all iodine spices, the resuspension of elemental iodine will produce the same thyroid organ dose irrespective of the form of iodine.

Resuspension rate of elemental iodine (sec")

= 2.32 (+/-2.00) x 10 e-"0 °F = 4.32 x 10" e"6O"r Resuspension rate of elemental iodine 7, (hrf1)

= 4.32 x 3600 x 10"5 e"600T The resuspension rates of elemental iodine at various drywell temperatures are calculated in Table 5.

The net deposition of elemental iodine on the pipe surface is the difference of deposition rate and resuspension rate. The net elemental iodine deposition rates at various drywell temperatures are calculated in Tables 6 and 7 for the MSIV failed and intact steam lines respectively.

Net Deposition Rate of Elemental Iodine XC = ,ed - Xer 1/DF = 1 - 'i = exp('Xe*t) (Ref 10.2, Equations 4 & 5, page 196)

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CALCULATION CONTINUATION SHEET SHEET 35 of 84 CALC. NO.: H-I-ZZ-MDC-1880

REFERENCE:

LCR H02-01 & DCP 80048085 G. Patel/NUCORE, ORIGINATOR, DATE REV: I1/27/02 K. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 11/29/02 Where DF = decontamination factor I = filter efficiency for elemental iodine Xe = elemental iodine removal rate (hi")

t = time (hr)

Therefore, Elemental Iodine Filter Efficiency = 1 - e"(X

• The values net elemental iodine deposition rates (Xe) are obtained from Tables 6 & 7 and the corresponding filter efficiencies at various drywell temperatures are calculated in Tables 8 & 9 for the MSIV failed and intact steam lines respectively. The conservative values are used for each time step in RADTRAD model rather than using average values for each time step.

The elemental iodine removal efficiencies at various drywell temperatures are used along with aerosol removal efficiency (Section 7.4.1) in the RADTRAD3.02 MSIV release model.

7.5 ESF Leak Rates The design basis ESF leakage is 1 gpm, which is doubled and converted into cfrn as follows:

1 gallon/min x 2 x 1/7.481 t/gallon = 0.2673 cfmn 10% of ESF leakage becomes airborne = 0.10 x 0.2673 = 0.02673 cfie 7.6 CR Direct Dose From Filter Shine CR Filter Shine Dose Rate = 7.754E-03 mRem/hr CR Operator Exposure Time = 1 x (24 hr) + 0.60 (96 hr- 24 hr) + 0.40 (720 hr- 96 hr)

= 24 hr + 0.60 (72 hr) + 0.40 (624 hr) = 316.8 hr Total CR Dose From Filter Shine

= 7.754E-03 mRem/hr x 1/1000 Rem/mnRem x 316.8 hr = 2.4565E-03 Rem 7.7 FRVS Vent & Recirc, and CR Charcoal/HEPA Filters Efficiencies HEPA Filter:

In-place penetration testing acceptance criteria for the safety related HEPA filters are as follows:

I Nuclear Common Revision 9 1I I ula omnRvso

FRVS Vent HEPA Filter - in-laboratory testing penetration < 0.05% (Ref. 10.6.1)

FRVS Recirc HEPA Filter - in-laboratory testing penetration < 0.05% (Ref. 10.6.10)

CREF HEPA Filter - in-laboratory testing penetration < 0.05% (Ref. 10.6.13)

GL 99-02 (Ref 10.3) requires a safety factor of at least 2 should be used to determine the filter efficiencies to be credited in the design basis accident.

Testing penetration (%) = (100% - ,)/safety factor = (100% - 11)/2 Where il = HEPA filter efficiency to be credited in the analysis 0.05% = (100% -,n)/2 0.1%= (100%-11) 1I = 100% - 0.1% = 99.9%

Conservatively, the HEPA filter efficiency of 99% is credited in the analysis Charcoal Filter:

In-place penetration testing acceptance criteria for the safety related Charcoal filters are as follows:

CREF Recirculation Charcoal Filter - in- laboratory testing methyl iodide penetration < 0.5% (Ref.

10.6.14)

Testing methyl iodide penetration (%) = (100% - ij)/safety factor = (100% - 11)/2 Where 11 = CREF charcoal filter efficiency to be credited in the analysis CFREF Charcoal Filter 0.5% = (100% - Ti)/2 1%= (100%- 1) 1I = 100% - 1%= 99%

FRVS Vent Charcoal Filter Elemental Iodine 1 = 90% (Ref. 10.47)

Organic Iodine i1 = 90% (Ref. 10.47)

Safety Grade Filter Efficiency Credited (%)

Filter Aerosol Elemental TOrganic I Nuclear Common Revision 9 1 evso Nula Cmo

I CALCULATION CONTINUATION SHEET SHEET 37 of 84 CALC. NO.: H- 1-ZZ-MDC- 1880

REFERENCE:

LCR H02-01 & DCP 80048085 G. Patel/UCORE, ORIGINATOR, DATE REV: 11127/02 1 M. Drucker/NUCORE, REVIEWERNERIFIER, DATE 11/29/02 FRVS Vent 99 90 90 FRVS Recirc 99 0 0 Control Room 99 99 99 7.8 Isotopic Activities Released To Environment The isotopic activities released to the environment at various post-LOCA time intervals are listed in Tables 18 through 22. These isotopic activities are obtained from the RADTRAD computer Runs HAST350CLOO.O0, HAST350ESFOO.O0, and HAST35OMSOO.O0. This information is used in the MIDAS computer code to assess dose profile during a design basis accident for Emergency Planning.

7.9 Extended Uprated Power Level Original Licensed Power Level = 3,293 MWt (Ref. 10.34)

Proposed Power Level Increase = 20%

Instrument Uncertainty = 2% (Ref. 10.10)

Extended Uprated Power Level = 3,293 MWt x 1.20 x 1.02 z 4,031 MWt I Nuclear Common Revision 9 I NcerCm oResin9I

CALCULATION CONTINUATION SHEET SHEET 38 of 84 CALC. NO.: H-I -ZZ-MDC-1880

REFERENCE:

LCR H02-01 & DCP 80048085 G. Patel/NUCORE, ORIGINATOR, DATE REV: 11/27/02 1 M. Drucker/NUCORE, REVIEWERNERIFIER, DATE 11/29/02 8.0 RESULTS

SUMMARY

The results of AST analyses are summarized in the following sections:

8.1 Licensing Basis Analysis The results of analyses, which establish licensing basis for the deletion of MSIVSS, increased MSIV, and CR unfiltered inleakage, are summarized in the following table:

Post-LOCA Post-LOCA TEDE Dose (Rem)

Activity Release Receptor Location Path Control Room EAB LPZ Containment Leakage 1.05E+00 3.731-01 1.62E-01 (5.5 hr)

ESF Leakage 1.25E+00 1.91E-01 9.79E,-02 (14.2 hr)

MSIV Leakage 2.13E+00 2.63E+00 4.56E-01 (9.3 hr)

Containment Purge O.OOE+O0 O.OOE+O0 O.OOE+O0 Containment Shine 0.00E+00 O.OOE+00 0.00E+00 External Cloud 0.O0E+00 O.OOE+00 0.00E+00 CR Filter Shine 2.46E-03* O.OOE+00 0.00E+00 Total 4.43E+00 3.19E+00 7.16E-01 Allowable TEDE Limit 5. E+00 2.5E+01 2.5E+01 RADTRAD Computer Run No.

Containment Leakage HEPU350CLOO HEPU35OCLOO HEPU350CLOO ESF Leakage IEPU350ESFOO HEPU350ESFOO HEPU350ESFOO MSIV Leakage HEPU350MSOO HEPU350MSOO HEPU350MSOO

• CR filter shine dose is bounding (see Section 6.4.4, Item 5).

Revision 9 Nuclear Common I Nuclear Revision 9 1

CALCULATION CONTINUATION SHEET SHEET 39 of 84 CALC. NO.: H-I-ZZ-MDC-1880

REFERENCE:

LCR H02-01 & DCP 80048085 G. Patel/NUCORE, ORIGINATOR, DATE REV: 11127/02 I M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 1129/02 8.2 V&V of RADTRAD V3.02 Code The comparison of results of RADTRAD3.02 and HABIT1.0 codes are shown in the following table:

Comparison of Control Room Doses - Licensing Basis Case Post-LOCA Control Room Dose (Rem) Dose Dose HABIT RADTRAD Variation ID Cont+ESF+MSIV Cont+MSIV ESF Total (%)

Thyroid 3.3634E-01 2.2593E-01 1.1570E-01 3.4163E-01 +1.57%

Whole 2.3027E-02 2.3392E-02 6.0348E-06 2.3398E-02 +1.61%

Body I I I I Comparison of Exclusion Area Boundary Doses - Licensing Basis Case Post-LOCA Exclusion Area Boundary Dose (Rem) Dose Dose HABIT RADTRAD Variation ID Cont+ESF+MSIV Cont+MSIV ESF Total (%)

Thyroid 1.2500E+02 9.1537E+01 3.3431E+01 1.2497E+02 -0.02%

Whole 1.3480E+00 12.283E+00 1.4366E-01 1.3720E+00 +1.78%

Body I I I Comparison of Low Population Zone Doses - Licensing Basis Case Post-LOCA Low Population Zone Dose (Rem) Dose Dose HABIT RADTRAD Variation ID Cont+ESF+MSIV Cont+MSIV ESF Total (%)

Thyroid 1.6820E+01 1.1796E+01 5.0355E+00 1.6832E+01 +0.07%

Whole 2.4120E-01 2.3056E-01 1.5273E-02 2.4583E-01 +1.92%

Body___I _

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CALCULATION CONTINUATION SHEET SHEET 40 of 84 CALC. NO.: H-I-ZZ-MDC-1880

REFERENCE:

LCR H02-01 & DCP 80048085 G. Patel/NUCORE, ORIGINATOR, DATE REV: 11/27/02 1 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE I V29/02

9.0 CONCLUSION

S CONCLUSIONS:

The Section 8.0 results of this analysis indicate that the following changes can be implemented using the AST and guidance in the Regulatory Guide 1.183:

a the decreased control room unfiltered inleakage to 350 cfm,

• the reduced ESF leakage from 10 gpm to 1 gpm, 0 the reduced FRVS vent charcoal filter efficiencies, 0 the FRVS vent heater deleted,

• the FRVS recirculation charcoal filtration deleted, and

• the increased thermal power level to 4,031 MWt for proposed power uprate Adherence to guidance in the RG 1.183 and use of the specific values and limits contained in the technical specifications and as-built post-accident performance of safety grade ESF functions provide the assurance of sufficient safety margin, including a margin to account for analysis uncertainties in the proposed uses of an AST and the associated facility modifications and changes to procedures. Procedure HC.RA-AP.ZZ-0051 (Q), Rev 1,"Leakage Reduction Program," should be revised to incorporate the new ESF leak rate limit of 1 gpm established in this analysis.

The verification & validation of RADTRAD3.02 computer code (Section 8.2) demonstrates that the RADTRAD3.02 code produces the identical results within +/- 2% margin of error compared to HABIT1.0 code for the same source terms, release mechanisms, and dose conversion factors. RADTRAD has been developed and tested by NRC in accordance with the requirements of ANSI/ANS-10.4-1987 in Reference 10.2, in Section 3, "Quality Assurance." In addition to the use of these programming standards, various program elements were tested and examined to insure program quality and ability to produce accurate and consistent results with HABIT 1.1 code.

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10.0 REFERENCES

10.1 U.S. NRC Regulatory Guide 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, July 2000.

10.2 S.L. Humphreys et al., "RADTRAD V3.02: A Simplified Model for Radionuclide Transport and Removal and Dose Estimation," NUREG/CR-6604, USNRC, April 1998.

10.3 USNRC, "Laboratory Testing of Nuclear-Grade Activated Charcoal," NRC Generic Letter 99-02, June 3, 1999.

10.4 10 CFR 50.67, "Accident Source Term."

10.5 Calculation No. H-1-ZZ-MDC-1879, Rev 1, Control Room & Technical Support Center x/Qs Using ARCON96 Code.

10.6 HCGS Technical Specifications:

10.6.1 Specification 4.6.5.3.1.c. 1, FRVS Vent HEPA Filter Testing Criterion 10.6.2 Not Used.

10.6.3 Specification 4.6.5.3.1.c.3, FRVS Vent HEPA/Charcoal Filter Flow Rate Testing Criterion 10.6.4 Specification 6.8.4.f, Primary Containment Leak Rate Testing Program 10.6.5 Bases %,.6.2,Depressurization Systems 10.6.6 Specification 5.2.1, Containment Configuration 10.6.7 Specification 5.2.3, Secondary Containment 10.6.8 Specification 4.6.5.1, Secondary Containment Integrity 10.6.9 Specification 1.35, Rated Thermal Power.

10.6.10 Specification 4.6.5.3.2.c. 1, FRVS Recirc HEPA Filter Testing Criterion 10.6.11 Not used 10.6.12 Specification 4.6.5.3.2.c.3, FRVS Recirc HEPA/Charcoal Filter Flow Rate Testing Criterion I Nuclear Common Revision 9 I I ula omnRvso

10.6.13 Specification 4.7.2.c. 1,Control Room Emergency Filtration System Surveillance Requirements 10.6.14 Specification 4.7.2.c.2, Control Room Emergency Filtration System Surveillance Requirements 10.6.15 Specification 4.7.2.c.3, Control Room Emergency Filtration System Surveillance Requirements 10.6.16 Specification 4.7.2.e.3, Control Room Emergency Filtration System Surveillance Requirements 10.6.17 Specification 3.6.1.2.c, Primary Containment Leakage Limiting Condition For Operation 10.6.18 Specification 3.6.1.8, Drywell and Suppression Chamber Purge System 10.6.19 HCGS Technical Specification Table 3.6.3-1, Primary Containment Isolation Valves.

10.7 Federal Guidance Report 11, EPA-520/1-88-020, Environmental Protection Agency.

10.8 Federal Guidance Report 12, EPA-402-R-93-081, Environmental Protection Agency.

10.9 MicroShield Computer Code, V&V Version 5.05, Grove Engineering and A-0-ZZ-MCS-0209, Sheet 1, Rev 0, MicroShield 5.05.

10.10 U.S. NRC Regulatory Guide 1.49, Rev 1, Power Levels of Nuclear Power Plants.

10.11 Drawing No. 1-P-AB-01, Rev 18, System Isometric / Turbine Building Main Steam Lead.

10.12 Fabrication Isometric Main Steam Lead - Turbine Building Unit #1 Drawings:

a. 1-P-AB-001, Rev 11

b. 1-P-AB-002, Rev 9 C. 1-P-AB-003, Rev 9

d. 1-P-AB-004, Rev 9

e. I-P-AB-011, Rev 11 10.13 Piping Area Drawings:

a. P-1703-1, Rev 3, Reactor Building Area 17, Plan EL 100'-2".

b. P-1704-1, Rev 2, Reactor Building Area 17, Plan EL 112'-0".

c. P-1705-1, Rev 2, Reactor Building Area 17, Plan EL 121'-7-1/2".

d. P-1712-1, Rev 2, Reactor Building Area 17, Section B17 - B17.

e. P-1713-1, Rev 4, Reactor Building Area 17, Section C17 - C17.

Nuclear Common Revision 9 ]

CALCULATION CONTINUATION SHEET SHEET 43 of 84 CALC. NO.: H-I-ZZ-MDC-1880

REFERENCE:

LCR H02-01 & DCP 80048085 G. Patel/NUCORE, ORIGINATOR, DATE REV: 11V27/02 1 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 1i29/02

f. P-1403-1, Rev 2, Reactor Building Area 14, Plan At EL 102'-0".

g. P-1414-1, Rev 1, Reactor Building Area 14, Section D14 - D14.

h. Not Used

i. P-171 1-1, Rev 2, Reactor Building Area 17, Section A17 - A17.

10.14 Piping Class Sheet Drawing No. 10855-P-0500::

10.14.a Sheet 16, Rev 9, Class DBB 10.14.b Sheet 17, Rev 7, Class DBC 10.14.c Sheet 24, Rev 7, Class DLA 10.15 GE-NE-T2300759-00-02, HCGS Containment Analysis With 100 OF SACS Temperature, September 1998 (VTD 323835, Sheet 2, Rev 1).

10.16 Calculation No. 12-0025, Rev 3, "Drywell Volume & Torus Air & Water Volumes."

10.17 Specification 10855-M-786 (Q), Rev 11, Technical Specification For HVAC Air Filter Systems, Seismic Category I For The Hope Creek Generating Station.

10.18 Procedure HC.RA-AP.ZZ-0051(Q), Rev. 1, Leakage Reduction Program 10.19 Calculation No. GU-0013, Rev. 4, Filtration Recirculation and Ventilation System Exhaust Rate 10.20 Drawing M-76-1, Rev. 18, P&ID Reactor Building Air Flow Diagram 10.21 CR961030231 Act. 0010 Response, Secondary Containment 10.22 NRC Report AEB-98-03, "Assessment of Radiological Consequences For the Perry Pilot Plant Application Using the Revised (NUREG-1465) Source Tenn.

10.23 MSIV Leakage Iodine Transport Analysis By J.E. Cline & Associates, March 26, 1991, Contract NRC-03-87-029, Task Order 75 10.24 NUREG/CR-2713, Vapor Deposition Velocity Measurements and Correlations for 12 and CsI, May 1982.

10.25 Calculation No. No. H-i -ZZ-MDC-0364, Rev 0, Drywell Temperature After Recirculation Line Break.

10.26 EQE International, Inc., Report No. 200235-R-01, November 12, 1998, Hope Creek Nuclear Plant Main Steam Isolation System Alternate Leakage Treatment Pathway Seismic Evaluation.

10.27 General Arrangement Drawings:

I Nuclear Common Revision 9 1 Nula omo eiso

CALCULATION CONTINUATION SHEET SHEET 44 of 84 CALC. NO.: H-I-ZZ-MDC-1880

REFERENCE:

LCR H02-01 & DCP 80048085 G. Patel/NUCORE, ORIGINATOR, DATE REV: 11/27/02 1 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 11/29/02 10.27.a P-0006-0, Rev 7, Plan EL 153'-0" and 162'-0" 10.27.b P-001 1-0, Rev 5, Sections C-C & D-D 10.28 Equipment Location Drawings:

10.28.a P-0035-0, Rev 10, Service & Radwaste Area Plan EL 137'-0" 10.28.b P-0036-0, Rev 16, Service & Radwaste Area Plan EL 153'-0"& 155'-3" 10.28.c P-0055-0, Rev 15, Control & D/G Area, Plan EL 137'-0" & EL 146'-0" & EL 150'-0" 10.28.d P-0056-0, Rev 16, Control & D/G Area, Plan EL 155'-3" & EL 163'-6" 10.29 Auxiliary Bldg - Control Area Drawings:

10.29.a C-1317-0, Rev 22, Floor Plan EL 155'-3" Area 25 10.29.b C-1319-0, Rev 12, Floor Plan EL 155'-3" Area 26 10.29.c C-1321-0, Rev 5, Roof Plan EL 172-0" Area 25 10.29.d C-1323-0, Rev 4, Roof Plan EL 172-0" Area 26 10.30 Auxiliary Bldg - Control Area Drawings:

10.30.a C-1313-0, Rev 11, Floor Plan EL 137'-0" Area 25 10.30.b C-1315-0, SH 2, Rev 3, Floor Plan EL 137'-0" Area 26 10.31 Auxiliary Bldg - Diesel Generator Area Drawings:

10.3l.a C-1413-0, Rev 20, Floor Plan EL 146'-0", EL 150'-0", EL 155'-3" Area 27 10.3l.b C-1415-0, Rev 22, Floor Plan EL 146'-0", EL 150'-0", EL 155'-3" Area 28 10.32 Calculation No. H-1-ZZ-MDC-1820, Rev 0, Offsite Atmospheric Dispersion Factors.

10.33 Calculation No. H-1-ZZ-MDC-1882, Rev 0, Control Room Envelope Volume.

10.34 NRC Safety Evaluation Report NUREG-1048, October 1984, Operation of Hope Creek Generating Station.

10.35 Drawing No. C-0738-0, Rev 6, Reactor Building Dome Reinforcement Plan Section & Details.

10.36 NRC Safety Evaluation for Amendment No. 30.

I Nuclear Common Revision 9 I Nula omnRvso

CALCULATION CONTINUATION SHEET SHEET 45 of 84 CALC. NO.: H-1-ZZ-MDC-1880

REFERENCE:

LCR H02-01 & DCP 80048085 G. Patel/NUCORE, ORIGINATOR, DATE REV: 11/27/02 1 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 11/29/02 10.37 Specification No 10855-P-0501, Rev 34, Line Index For The Hope Creek Generating Station.

10.38 American Air Filter Drawing No. M786(Q)-5(1), Rev 10, Housing Assy Filter (Control Room Emergency Filter).

10.39 HVAC Area Drawings:

10.39.a P-9266-1, Rev 25, Aux Bldg Area 26, Plan At EL 155'-3"& 163'-6" 10.39.b P-9256-1, Rev 24, Aux Bldg Area 25, Plan At EL 155'-3"& 175'-0" 10.39.c P-9267-1, Sheet 1 of 4, Rev 17, Aux Building Area 25 & 26 Sections 10.40 U.S. NRC Standard Review Plan 6.4, Control Room Habitability System.

10.41 P&ID M-57-1, Rev 36, Containment Atmosphere Control.

10.42 P&ID M-78-1, Rev 9, Containment Hydrogen Recombination System.

10.43 Calculation No. H- 1-ZZ-MDC- 1866, Rev 0, Hope Creek Post-Accident pH.

10.44 Order No.80028003, Confirmatory Inputs For H-1-ZZ-MDC-1880 10.45 GE-NE-0000-0008-3534-01, DRF 0000-0004-6923, Revision 0, Class III, October 2002, Project Task Report T0802, Radioactive Source Term - Core Inventory 10.46 PSE&G Vendor Technical Document (VTD) No. 325236, Control Room Envelop Inleakage Testing At Hope Creek Generating Station, Final Report, 2001.

10.47 E-mail From John P. Cichello To Gopal Patel, Dated 03/15/02,

Subject:

FRVS Vent Charcoal Filter Efficiencies (Attachment A) 10.48 Critical Software Package Identification No. A-0-ZZ-MCS-0225, Rev 0, RADTRAD Computer Code.

10.49 DCP No. 80042273, Rev 0, FRVS Charcoal Relaxation.

10.50 Not Used.

10.51 Draft Safety Evaluation Report, November 14, 2000, Increase of Allowable Main Steam Isolation Valve (MSIV) leak rate and Deletion of MSIV Sealing System (TAC No. MA9978).

10.52 HCGS Procedure No. HC.RA-IS.ZZ-0010(Q), Rev 8, Containment Isolation Valve Type C Leak Rate Test.

I Nuclear Common Revision 9 1 omn eiin9I Nula

CALCULATION CONTINUATION SHEET SHEET 46 of 84 CALC. NO.: H-1-ZZ-MDC-1880

REFERENCE:

LCR H02-01 & DCP 80048085 G. Patel/NUCORE, ORIGINATOR, DATE REV: 11t27/02 I M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 11/29/02 10.53 E-mail Response From William T. Kittle To Gopal J. Patel, Dated 03/15/02,

Subject:

MSIV Type C Leak Rate Test (Attachment B).

10.54 NRC letter to PSEG Nuclear dated October 3, 2001, "Hope Creek Generating Station - Issuance of Amendment Re: Increase In Allowable Main Steam Isolation Valve (MSIV) Leakage Rate and Elimination of MSIV Sealing System (TAC No. MB1970)."

10.55 Calculation No. H-1-KE-MDC-1 898, Rev 0, Radiation Consequences of Removing Reactor Well Shield Before Cold Shutdown.

I Nuclear Common Revision 9 1 eiso  !

Nula omo

CALCULATION CONTINUATION SHEET SHEET 47 of 84 CALC. NO.: H-I-ZZ-MDC-1880

REFERENCE:

LCR H02-01 & DCP 80048085 G. Patel/NUCORE, ORIGINATOR, DATE REV: 11/27/02 1 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 11/29/02 11.0 TABLES Table 1A Unrated Core Inventory Including Parent/Daughter IsotoDes Per RADTRAD Table 1.4.3.3-2 Core Core Core Core Isotope Inventory Inventory Isotope Inventory Inventory (Ci/MWO (Ci/MWO (P/MW0) (C/m~t)

CO-58* 1.529E+02 1.529E+02 CO-60* 1.830E+02 1.830E+02 KR-85 4.711E+02 4.711E+02 TE-132 3.917E+04 3.917E+04 KR-85M 5.908E+03 5.908E+03 1-131 2.779E+04 2.779E+04 KR-87 1.097E+04 1.097E+04 1-132 3.991E+04 3.991E+04 KR-88 1.539E+04 1.539E+04 1-133 5.454E+04 5.454E+04 RB-86 1.300E+02 1.300E+02 1-134 5.937E+04 5.937E+04 SR-89 2.056E+04 2.056E+04 SR-90 3.790E+03 3.790E+03 R-5 ME+ XE-133 5.306E+04 5.306E+04 XE-135 1.482E+04 1.482E+04 SR-92 2.990E+04 2.9901+04 CS-134 1.319E+04 1.391E+04

-Y-90! 3.9813E+03 3.981E+03 CS-136 3.704E+03 3.704E+03 Y-92 3.0053E+04 3.005E+04 U3 I 33 Y-93 3.6073E+04 3.6073E+04 BA-139 4.760E+04 4.760E+04 S LA-140 4.981E+04 4.98]E+04 T LA-141 4.325E+04 4.325E+04 4.1!3413+04 LA-142 4.134E+04 NB-95 42371E+04 4.2373E+04 CE-141 4.350E+04 4.350E+04 MO-99 52781E+04 5.2781E+04 CE-143 3.9101E+04 53.9101E+04 4PND-147 1.783E+04 1.78313+04 NP-239 6.91713+05 6.91713+05 RH-105 3.23713+04 3.23713+04 PU-238 3.44213+02 3.44213+02 SB-127 3.379E+03 3.37913+03 PU-239 1.33313+01 1.33313+01 SB-129 9.56913+03 9.56913+03 PU-240 2.67513+01 2.67513+01 TE-127 3.35513+03 3.35513+03 PU-241 5.41913+03 5AI9E3+03 TE-127M 4.50813+02 4.50813+02 AM-241 7.26613+00 7*26613+00 TE-129 9.43013+03 9.43013+03 CM-242 2.56713+03 2.567E+03 TE- 129M 1.4011E+03 1.4011E+03 CM-244 5.18813+02 5.18813+02

• CO-58 & CO-60 activities are obtained from RADTRAD User's Manual, Table 1.4.3.2-3 (Ref. 10.2)

INuclear Common Revision 9

CALCULATION CONTINUATION SHEET SHEET 48 of 84 CALC. NO.: H-I-ZZ-MDC-1880

REFERENCE:

LCR H02-01 & DCP 80048085 G. Patel/NUCORE, ORIGINATOR, DATE REV: 11/27/02 1 M. Drucker/NUCORE, REVIEWERNVERIFIER, DATE 11/29/02 Table 1B Uprated Core Inventory Total Total Core Core Isotope Inventory Isotope Inventory (Ci/MWt) (Ci/MWt)

CO-58* 1.529E+02 TE-131M 2.876E+04 CO-60* 1.830E1+02 TE-132 3.917E+04 KR-85 4.71 !E+02 1-131 2.779E+04 KR-85M 5.908E+03 1-132 3.991E+04 KR-87 1.097E+04 1-133 5.454E+04 KR-88 1.539E+04 1-134 5.937E+04 RB-86 1.300E+02 1-135 6.235E+04 SR-89 2.056E+04 XE-1 33 5.306E+04 SR-90 3.790E+03 XE-135 1.482E+04 SR-91 4.231E+04 CS-134 1.319E+04 SR-92 2.990E3+04 CS-136 3.704E+03 Y-90 3.981E+03 CS-137 1.096E+04 Y-91 2.750E+04 BA-139 4.760E+04 Y-92 3.005E+04 BA-140 4.590E+04 Y-93 3.607E+04 LA-140 4.98 IE+04 ZR-95 4.217E3+04 LA-141 4.325E+04 ZR-97 1.307E+05 LA-142 4.134E+04 NB-95 4.237E+04 CE-141 4.350E+04 MO-99 5.278E+04 CE-143 3.910E+04 TC-99M 4.621E+04 CE-144 7.234E+04 RU- 103 8.941E+04 PR-143 3.783E+04 RU- 105 3.529E+04 ND-147 1.783E+04 RU-106 4.722E+04 NP-239 6.917E+05 RH-105 3.237E+04 PU-238 3.442E+02 SB-127 3.379E+03 PU-239 1.333E+01 SB-129 9.569E+03 PU-240 2.675E+01 TE-127M 4.508E+02 PU-241 5.419E+03 TE- 127 3.355E+03 AM-241 7.266E+00 TE-129M 1.401 E+03 CM-242 2.576E+03 TE-129 9.430E+03 CM-244 5.188E+02 it I Nuclear Common Revision 99 1I Revision I Nuclear Common

CALCULATION CONTINUATION SHEET SHEET 49 of 84 CALC. NO.: H-I-ZZ-MDC-1880

REFERENCE:

LCR H02-01 & DCP 80048085 G. Patel/NUCORE,II ORIGINATOR, DATE REV: 11/27/02 1 M. Drucker/NUCORE, REVIEWERNERIFIER, DATE 11/29/02 Table 2 Elemental Iodine Deposition Velocity - MSIV Leakage Time Temp Temp Deposition Deposition Degree* Degree (2809/7) -12.5 Velocity Velocity F K cm/sec m/sec 0 340 444.26 -6.18 0.002076 2.076E-05 3 320 433.15 -6.01 0.002442 2.442E-05 6 250 394.26 -5.38 0.004630 4.630E-05 24 208 370.93 -4.93 0.007248 7.248E-05 96 180 355.37 -4.60 0.010096 1.010E-04 240 170 349.82 -4.47 0.011446 1.145E-04 480 150 338.71 -4.21 0.014896 1A90E-04 720 1 1

• From Design Input 5.3.1.8, Table 6 I Nuclear Common Revision 9 1 I Nula omnRvso

CALCULATION CONTINUATION SHEET SHEET 50 of 84 CALC. NO.: H-I-ZZ-MDC-1880

REFERENCE:

LCR H02-01 & DCP 80048085 G. Patel/NUCORE, ORIGINATOR, DATE REV: 11/27/02 1 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 11/29/02 Table 3 Elemental Iodine Deposition Rate - MSIV Failed Line Deposition Main Steam Line Elemental Velocity Total Total Iodine Time Surface Volume Removal Area Rate 2 3 (h)

Hr m/sec (m ) (m )

A B C (AXB) x3600/C 0 2.076E3-05 247.77 39.58 0.4679 3 2.442E-05 247.77 39.58 0.5503 6 4.630E-05 247.77 39.58 1.0433 24 7248E-05 247.77 39.58 1.6333 96 1.010E-04 247.77 39.58 2.2752 240 1.145E-04 247.77 39.58 2.5796 480 1.490E-04 247.77 39.58 3.3570 720 1 1 1 1 A From Table 2 B & C From Section 7.2.5 Revision I Nuclear Nuclear Common Revision 9 i

CALCULATION CONTINUATION SHEET SHEET 51 of 84 CALC. NO.: H-I-ZZ-MDC-1880

REFERENCE:

LCR H02-01 & DCP 80048085 G. Patel/NUCORE, ORIGINATOR, DATE REV: 11/27/02 1 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 11/29/02 Table 4 Elemental Iodine Deposition Rate - MSIV Intact Lines Deposition Main Steam Line Elemental Velocity Total Total Iodine Time Surface Volume Removal Area Rate (hr 1)

Hr m/sec (mi) (mn)

A* B C (AXB) x3600/C 0 2.0761-05 262.46 41.83 0.4690 3 2.442E-05 262.46 41.83 0.5516 6 4.630E-05 262.46 41.83 1.0457 24 7.248E-05 262.46 41.83 1.6371 96 1.010E-04 262.46 41.83 22.805 240 1.145E-04 262.46 41.83 2.5855 480 1A90E-04 262.46 41.83 3.3647 720 1 1 1 1 A From Table 2 B & C From Section 7.2.6 evso Nula Cmo I Nuclear Common Revision 9 1

CALCULATION CONTINUATION SHEET SHEET 52 of 84 CALC. NO.: H-I-ZZ-MDC-1880

REFERENCE:

LCR H02-01 & DCP 80048085 G. Patel/NUCORE, ORIGINATOR, DATE REV: 11/27/02 1 M. Drucker/NUCORE, REVIEWERNVERIFIER, DATE 11/29/02 Table 5 Elemental Iodine Resuspension Rate - MSIV Leakage Post-LOCA Temp Temp Resuspension Time Degree Degree -600/T Rate (hr) F K (he'A) 0 340 444.26 -1.35 0.0403 3 320 433.15 -1.39 0.0389 6 250 394.26 -1.52 0.0340 24 208 370.93 -1.62 0.0309 96 180 355.37 -1.69 0.0287 240 170 349.82 -1.72 0.0280 480 150 338.71 -1.77 0.0265 720 1 1 1 1 0 6 00 Resuspension Rate (sec)" f 2.32 (2.00) x 10"s e* 0I' = 4.32 x 10- eC /T Resuspension Rate (hr)' = 4.32 x 3600 x 1e e" 00 /T I Nuclear Common Revision 9 1 eiin9I Nula omn

CALCULATION CONTINUATION SHEET SHEET 53 of 84 CALC. NO.: H-I-ZZ-MDC-1880

REFERENCE:

LCR H02-01 & DCP 80048085 G. Pate/NUCORE, ORIGINATOR, DATE REV: 11!27/02 1 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 11/29/02 Table 6 Net Elemental Iodine Removal Rate - MSIV Failed Line Iodine Iodine Net Iodine Post-LOCA Temp Removal Resuspension Removal Time Degree Rate Rate Rate A B X=A-B (hr) F (hr-1) (hr-1) (hr-I) 0 340 0.4679 0.0403 0.4276 3 320 0.5503 0.0389 0.5114 6 250 1.0433 0.0340 1.0094 24 208 1.6333 0.0309 1.6024 96 180 2.2752 0.0287 2.2465 240 170 2.5796 0.0280 2.5516 480 150 3.3570 0.0265 3.3305 720 1 A From Table 3 B From Table 5 eiso

[Icea omo I Nuclear Common Revision 9 1

CALCULATION CONTINUATION SHEET SHEET 54 of 84 CALC. NO.: H-I-ZZ-MDC-1880

REFERENCE:

LCR H02-01 & DCP 80048085 G. Patel/NUCORE, ORIGINATOR, DATE REV: 11/27/02 1 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 11/29/02 Table 7 Net Elemental Iodine Removal Rate - Intact Lines Iodine Iodine Net Iodine Post-LOCA Temp Removal Resuspension Removal Time Degree Rate Rate Rate A B 7ffA-B (hr) F (hr-I) (hr-1) (hr-l) 0 340 0.4690 0.0403 0.4287 3 320 0.5516 0.0389 0.5127 6 250 1.0457 0.0340 1.0118 24 208 1.6371 0.0309 1.6062 96 180 2.2805 0.0287 2.2518 240 170 2.5855 0.0280 2.5575 480 150 3.3647 0.0265 3.3383 720 1 1 1 A From Table 4 B From Table 5 Revision 9 I Nuclear Nuclear Common Common Revision 9 1

CALCULATION CONTINUATION SHEET SHEET 55 of 84 CALC. NO.: H-1-ZZ-MDC-1880

REFERENCE:

LCR H02-01 & DCP 80048085 G. Patel/NUCORE, ORIGINATOR, DATE REV: 11/27/02 1 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 11/29/02 Table 8 Elemental Iodine Removal Efficiency - MSIV Failed Line Net Elemental Post-LOCA Temp Iodine Iodine Time Degree Removal Removal Rate Efficiency XB (hr) F (hr-l) (%)

0 340 0.4276 34.79 3 320 0.5114 40.03 6 250 1.0094 63.56 24 208 1.6024 79.86 96 180 2.2465 89.42 240 170 2.5516 92.20 480 150 3.3305 96.42 720 XfFrom Table 6 B= l-ex Common Revision 9 I I Nuclear Nudear Common Revision ýi

CALCULATION CONTINUATION SHEET SHEET 56 of 84 CALC. NO.: H-1-ZZ-MDC-1880

REFERENCE:

LCR H02-01 & DCP 80048085 G. Patel/NUCORE, ORIGINATOR, DATE REV: 11/27/02 1 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 11/29/02 Table 9 Elemental Iodine Removal Efficiency - Intact Lines Net Elemental Post-LOCA Temp Iodine Iodine Time Degree Removal Removal Rate Efficiency

&. B (hr) F (hr-i) (%)

0 340 0.4287 34.87 3 320 0.5127 40.11 6 250 1.0118 63.64 24 208 1.6062 79.94 96 180 22518 89.48 240 170 2.5575 92.25 480 150 3.3383 96.45 720 A From Table 7 B = I-e"'t Revision 9 I Nuclear Nuclear Common Common Revision 9 1

CALCULATION CONTINUATION SHEET SHEET 57 of 84 CALC. NO.: H-I -ZZ-MDC- 1880

REFERENCE:

LCR H02-01 & DCP 80048085 G. Patel/NUCORE, ORIGINATOR, DATE REV: 11/27/02 1 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 11/29/02 Table 10 Post-LOCA Containment Leakage Activity in CR With Charcoal/HEPA Filters (c)

Isotope 0-0.33 0.33-0.5 0.5-2 2-4 4-8 8-24 24-96 96-720 Total Co-58 0.OOE+0O 0.OOE+00 2.94E-08 7.97E-09 3.00E-10 0.OOE+00 0.OOE+00 0.OOE+00 3.76E-08 Co-60 0.OOE+00 0.OOE+00 3.52E-08 9.55E-09 3.61E-10 5.75E-14 0.OOE+00 O.OOE+00 4.5 1E-08 Kr-85 3.31E-05 4.77E-05 1.21E-02 2.3 1E-02 4.82E-02 3.67E-02 1.43E-02 9.79E-03 1.44E-01 Kr-85m 1.14E-03 1.61E-03 3.23E-01 4.52E-01 5.08E-01 3.25E-02 1.84E-07 0.OOE+00 1.32E+00 Kr-87 1.83E-03 2.40E-03 2.69E-01 1.73E-01 4.07E-02 5.05E-06 0.OOE+00 0.OOE+00 4.87E-01 Kr-88 2.73E-03 3.77E-03 6.63E-01 7.77E-01 6.10E-01 9.35E-03 0.OOE+00 0.00E+00 2.07E+00 Rb-86 1.48E-06 1.40E-06 3.08E-07 6.99E-08 2.60E-09 O.OOE+00 0.OOE+00 0.OOE+00 36.6E-06 Sr-89 0.OOE+00 0.OOE+00 4.26E-05 1.16E-05 4.35E-07 6.88E-11 5.94E-12 2.69E-12 5A66E-05 Sr-90 0.00E+00 O.OOE+00 3.02E-06 8.20E-07 3.09E-08 4.94E-12 4.44E-13 2.87E-13 3.87E-06 Sr-91 0.OOE+00 O.OOE+00 4.79E-05 1.12E-05 3.17E-07 0.00E+00 0.00E+00 O.OOE+00 5.95E-05 Sr-92 0.00E+00 0.OOE+00 3.47E-05 5.65E-06 7.67E-08 O.OOE+0O 0.OOE+00 0.00E+00 4.04E-05 Y-90 0.00E+00 0.OOE+00 3.16E-08 8.41E-09 3.04E-10 0.OOE+0O 0.OOE+00 O.OOE+00 4.03E-08 Y-91 0.00E+00 O.OOE+00 5.20E-07 1.41E-07 5.32E-09 8.42E-13 O.OOE+00 0.00E+00 6.67E-07 Y-92 0.001E+00 0.OOE+00 3.93E-07 7.21E-08 1.24E-09 0.OOE+00 O.OOE+00 0.00E+00 4.66E-07 Y-93 0.OOE+00 0.OOE+00 5.76E-07 1.36E-07 3.911E-09 0.OOE+00 0.OOE+00 0.00E+00 7.17E-07 Zr-95 0.OOE+00 0.OOE+00 6.85E-07 1.86E-07 7.00E-09 1.111E-12 0.00E+3 0.0OOE+00 8.78E-07 Zr-97 0.OOE+00 0.00E+00 6.50E-07 1.63E-07 521E-09 O.OOE+00 0.OOE+00 0.OOE+00 8.188E-07 Nb-95 0.OOE+00 0.OOE+00 6.47E-07 1.76E-07 6.60E-09 1.04E-12 0.OOE+00 0.OOE+00 8.29E-07 Mo-99 0.00E+00 0.00E+00 9.15E-06 2.43E-06 8.81E-08 1.19E-1 1 0.00E+00 0.00E+00 1.17E-05 Tc-99m O.OOE+00 0.OOE+00 6.41E-06 1.38E-06 3.29E-08 0.OOE+00 0.OOE+00 0.00E+00 7.82E-06 Ru-103 0.00E+0O 0.OOE+00 7.09E-06 1.92E-06 7.23E-08 1.14E-1 I 9.72E-13 0.00E+00 9.08E-06 Ru-105 0.00E+0 0.0OOE+00 3.46E-06 6.87E-07 1.39E-08 O.OOE+00 0.00E+00 O.OOE+00 4.16E-06 Ru-106 0.OOE-0 0.0OOE+00 1.93E-06 5.23E-07 1.97E-08 3.14E-12 2.81E-13 1.74E-13 2.47E-06 Rh-105 0.00E+00 O.OOE+00 3.39E-06 8.86E-07 3.09E-08 0.OOE+00 0.OOE+00 0.00E+00 4.3 1E-06 Sb-127 0.001E+00 0.OOE+00 8.80E-06 2.36E-06 8.63E-08 1.22E-11 1.22E-II 0.OOE+00 1.12E-05 Sb-129 0.OOE+00 0.OOE+00 2.25E-05 4.43E-06 8.81E-08 0.00E+00 0.00E+00 0.OOE+00 2.70E-05 Te-127 O.OOE+00 0.OOE+00 7.46E-06 1.75E-06 4.90E-08 0.OOE+00 0.00E+00 0.OOE+00 9.25E-06 Te-127m 0.OOE+00 0.OOE+00 1.16E-06 3.16E-07 1.19E-08 1.89E-12 2.10E-13 0.OOE+00 1.49E-06 Te-129 0.00E+00 0.OOE+00 8.81E-06 7.24E-07 2.50E-09 0.OOE+00 0.OOE+00 0.OOE+00 9.54E-06 Te-129m 0.OOE+00 0.00E+00 7.64E-06 2.07E-06 7.79E-08 1.23E- II 1.04E-12 0.OOE+00 9.79E-06 I Nuclear Common Revision 9 1

! Nula omnRvso

CALCULATION CONTINUATION SHEET SHEET 58 of 84 CALC. NO.: H-1-ZZ-MDC-1880

REFERENCE:

LCR H02-01 & DCP 80048085 G. Patel/NUCORE, ORIGINATOR, DATE REV: 11t27/02 1 M. Drucker/NUCORE, REVIEWERNERIFIER, DATE 1!/29/02 Table 10 (Cont'd)

Post-LOCA Containment Leakage Activity in CR With Charcoal/HEPA Filters (C)

Isotope 0-0.333 0.33-0.5 0.5-2 2-4 4-8 8-24 24-96 96-720 Total Te-131m 0.00E+00 0.00E+00 1.40E-05 3.64E-06 1.25E-07 0.00E+00 0.00E+00 0.00E+00 1.78E-05 Te-132 O.OOE+00 0.00E+00 1.41E-04 3.76E-05 1.37E-06 1.90E-10 9.02E-12 0.00E+00 1.80E-04 1-131 2.72E-03 2.58E-03 6.72E-04 1.75E-04 3.15E-05 9.83E-06 2.69E-06 1.97E-07 6.19E-03 1-132 3.62E-03 3.26E-03 5A5E-04 7.83E-05 4.27E-06 1.14E-08 0.OOE+00 0.00E+00 7.5 1E-03 1-133 5.65E-03 5.33E-03 1.33E-03 3.27E-04 5.21E-05 1.01E-05 3.25E-07 0.00E+00 1.27E-02 1-134 4.81E-03 3.99E-03 3.20E-04 1.73E-05 1.33E-07 0.00E+00 0.00E+00 0.00E+00 9.14E-03 1-135 5.20E-03 4.84E-03 1.08E-03 2.31_E-04 2.77E-05 1.71E-06 3.18E-10 0.00E+00 1.14E-02 Xe-133 7.16E-03 1.03E-02 2.59E+00 4.89E1+00 9.98E+00 6.96E+00 1.82E+00 4.04E-02 2.63E+01 Xe-135 1.66E-03 2.36E-03 5.35E-01 8.76E-01 1.35E+00 3.03E-01 4.86E-04 0.00E+00 3.06E+00 Cs-134 4.46E-04 43.3E-04 9.33E-05 2.12E-05 7.92E-07 1.26E-10 1.13E-11 7.16E-12 9.84E-04 Cs-136 1.20E-04 1.13E-04 2.49E-05 5.64E-06 2.09E-07 3.21E-11 2.47E-12 0.00E+00 2.64E-04 Cs-137 2.67E-04 2.53E-04 5.58E-05 1.2711-05 4.7413-07 7.56E-11 6.80E-12 4.40E-12 5.89E-04 Ba-139 O.OOE+00 0.00E+00 2.81E-05 2.79E-06 1.41E-08 0.OOE+00 0.00E+00 0.00E+00 3.09E-05 Ba-140 0.OOE+00 0.00E+00 7.54E-05 2.04E-05 7.63E-07 1.17E-10 8.96E-12 1.78E-12 9.66E-05 La-140 0.OOE+00 0.00E+00 7.47E-07 1.96E-07 6.911E-09 0.00E+00 0.OOE+00 0.00E+00 9.50E-07 La-141 0.00E+00 0.00E+00 5.02E-07 9.58E-08 1.79E-09 0.00E+00 0.00E+00 0.00E+00 5.99E-07 La-142 0.00E+00 0.00E+00 2.79E-07 3.09E-08 1.75E-08 0.00E+00 0.00E+00 0.00E+00 3.28E-07 Ce-141 0.00E+00 0.00E+00 1.72E-06 4.65E-07 1.75E-08 2.75E-12 1.62E-13 0.00E+00 2.20E-06 Ce-143 0.00E+00 0.003E+00 1.61E-06 4.18E-07 1.45E1-08 0.OOE+00 0.OOE+00 0.00E+00 2.04E-06 Ce-144 0.00E+00 0.00E+00 1.12E-06 3.03E-07 1.14E-08 1.82E-12 2.04E-13 9.88E-14 1.43E-06 Pr-143 0.OOE+00 0.00E+00 6.53E-07 i.77E-07 6.611E-09 0.00E+00 0.00E+00 0.00E+00 8.36E-07 Nd-147 0.00E+00 0.00E+00 2.91E-07 7.87E-08 2.94E-09 0.00E+00 0.00E+00 0.00E+00 3.73E-07 Np-239 0.00E+00 0.OOE+00 2.13E-05 5.64E-06 2.03E-07 2.66E-1 1 0.00E+00 0.00E+00 2.711E-05 Pu-238 0.OOE+00 0.00E+00 1.52E-09 4.12E-10 1.56E-i 1 2.48E-15 0.00E+00 0.00E+00 1.95E-09 Pu-239 0.OOE+00 0.00E+00 3.85E-10 1.05E-10 3.95E-12 6.29E-16 5.66E-17 3.67E-17 4.93E-10 Pu-240 0.00E+00 0.00E+00 4.82E-10 i.31E-10 4.94E-12 7.88E-16 7.08E-17 4.59E-17 6.18E-10 Pu-241 0.00E+00 0.00E+00 8.29E-08 2.25E-08 8.50E-10 1.36E-13 1.22E-14 7.87E-15 1.06E-07 Am-241 0.00E+00 0.OOE+00 3.37E-11 9.16E-12 3.46E-13 0.00E+00 0.00E+00 0.OOE+00 4.32E-1 1 Cm-242 0.00E+00 0.00E+00 8.90E-09 2.42E-09 9.12E-1 1 0.00E+00 0.00E+00 0.00E+00 1.14E-08 Cm-244 0.00E+00 0.00E+00 4.81E-10 1.31E-10 4.93E-12 0.00E+00 0.00E+00 0.00E+00 6.16E-10 From RADTRAD Computer Run HAST1000CL02 I Nuclear Common Revision 9 1

! Nula omnRvso

CALCULATION CONTINUATION SHEET SHEET 59 of 84 CALC. NO.: H-I-ZZ-MDC-1880

REFERENCE:

LCR H02-01 & DCP 80048085 G. Patel/NUCORE, ORIGINATOR, DATE REV: i 1/27/02 1 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 11/29/02 Table 11 Post-LOCA Containment Leakage Activity In CR Without Charcoal/HEPA Filters (Ci)

Isotope 0-0.333 0.33-0.5 0.5-2 2-4 4-8 8-24 24-96 96-720 Total Co-58 0.OOE+0O 0.OOE+00 9.30E-08 4.69E-08 2.1 SE-09 0.00E+00 0.OOE+00 O.OOE+00 1.42E-07 Co-60 0.OOE+00 0.OOE+00 1.11E-07 5.62E-08 2.61E-09 3.37E-13 2.34E-14 1.50E-14 1.70E-07 Kr-85 3.3 1E-05 4.77E-05 1.21E-02 2.31E-02 4.82E-02 3.67E-02 1.43E-02 9.79E-03 1.44E-01 Kr-85m 1.14E-03 i.61E-03 3.23E-01 4.52E-01 5.08E-01 3.25E-02 1.84E-07 0.OOE+00 1.32E+00 Kr-87 1.83E-03 2.40E-03 2.69E-01 1.73E-01 4.07E-02 5.05E-06 0.OOE+00 0.OOE+00 4.87E-01 Kr-88 2.73E-03 3.77E-03 6.63E-01 7.77E-01 6.10E-01 9.35E-03 0.00E+00 0.OOE+00 2.07E+00 Rb-86 1.48E-06 1.40E-06 1.17E-06 4.34E-07 1.89E-08 2.36E-12 0.00E+0 0.0OOE+00 4.50E-06 Sr-89 0.OOE+0 0.OOE+00 1.35E-04 6.80E-05 3.15E-06 4.03E-10 2.69E-11 1.22E-11 2.06E-04 Sr-90 O.OOE+00 0.OOE+00 9.56E-06 4.82E-06 2.24E-07 2.89E-1 1 2.01E-12 1.30E-12 1.46E-05 Sr-91 0.OOE+00 0.OOE+00 1.52E-04 6.61E-05 2.30E-06 9.21E- 1 0.OOE+0 0.OOE+00 21.0E-04 Sr-92 0.OOE+00 0.OOE+00 1.10E-04 3.32E-05 5.56E-07 0.00E+00 0.OOE+00 0.OOE+00 1.44E-04 Y-90 0.OOE+00 0.OOE+00 1.OOE-07 4.94E-08 2.20E-09 0.OOE+00 0.OOE+00 0.OOE+00 1.52E-07 Y-91 O.OOE+00 0.OOE+00 1.65E-06 8.30E-07 3.85E-08 4.93E-12 0.OOE+00 0.OOE+00 2.52E-06 Y-92 0.OOE+00 0.OOE+00 1.24E-06 4.24E-07 9.01E-09 0.OOE+00 0.00E+0 0.0OOE+00 1.68E-06 Y-93 0.OOE+00 0.OOE+00 1.82E-06 8.02E-07 2.83E-08 0.OOE+00 0.003E+0 0.OOE+00 2.65E-06 Zr-95 0.00E+0 0.0OOE+00 2.17E-06 1.09E-06 5.07E-08 6.49E-12 4.37E-13 0.OOE+00 3.3 1E-06 Zr-97 0.OOE+00 0.00E+00 2.06E-06 9.56E-07 3.77E-08 0.OOE+00 0.00E+00 0.OOE+00 3.05E-06 Nb-95 0.OOE+00 O.OOE+00 2.05E-06 1.03E-06 4.78E-08 6.09E-12 0.0OE+00 0.OOE+00 3.13E-06 Mo-99 0.OOE+00 0.00E+00 2.90E-05 1A3E-05 6.38E-07 6.95E-1 1 0.00E+00 0.OOE+00 4.39E-05 Tc-99m O.OOE+00 0.OOE+00 2.03E-05 8.12E-06 2.38E-07 0.00E+00 0.00E+00 0.OOE+00 2.86E-05 Ru-103 0.00E+00 0.OOE+00 2.24E-05 1.13E-05 5.24E-07 6.67E-11 44AOE-12 1.80E-12 3.42E-05 Ru-105 0.OOE+00 0.OOE+00 1.09E-05 4.04E-06 1.01E-07 0.OOE+00 0.OOE+00 0.OOE+00 1.51E-05 Ru-106 0.OOE+00 0.OOE+00 6.10E-06 3.07E-06 1.43E-07 1.84E-11 1.27E-12 7.86E-13 9.31E-06 Rh-105 0.OOE+00 0.OOE+00 1.07E-05 5.21E-06 2.24E-07 2.11E-11 O.OOE+00 0.OOE+00 1.62E-05 Sb-127 0.003E+00 0.OOE+00 2.79E-05 1.38E-05 6.25E-07 7.14E-1 I O.OOE+0O 0.OOE+00 4.23E-05 Sb-129 0.00E+0O 0.OOE+00 7.12E-05 2.61E-05 6.38E-07 0.OOE+00 0.OOE+00 0.OOE+00 9.79E-05 Te-127 0.OOE+0O 0.OOE+00 2.36E-05 1.03E-05 3.55E-07 0.OOE+00 0.OOE+00 0.OOE+00 3.42E-05 Te-127m 0.OOE+0O 0.OOE+00 3.68E-06 1.86E-06 8.63E-08 1.1IE-131 7.56E-13 4.15E-13 5.63E-06 Te-129 0.OOE+00 0.00E+00 2.79E-05 4.26E-06 1.81E-08 0.OOE+00 0.OOE+00 0.OOE+00 3.22E-05 Te-129m 0.00E+00 0.OOE+00 2.42E-05 1.22E-05 5.64E-07 7.17E- 11 4.69E-12 1.78E-12 3.69E-05 I Nuclear Common Revision 9 1 I Nula omnRvso

CALCULATION CONTINUATION SHEET SHEET 60 of 84 CALC. NO.: H-I-ZZ-MDC-1880

REFERENCE:

LCR H02-01 & DCP 80048085 G. Patel/NUCORE, ORIGINATOR, DATE REV: 11/27/02 1 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 11/29/02 Table 11 (Cont'd)

Post-LOCA Containment Leakage Activity in CR Without Charcoal/HEPA Filters (Ci)

Isotope 0-0.33 0.33-0.5 0.5-2 2-4 4-8 8-24 24-96 96-720 Total Te-131m 0.OOE+00 0.00E+00 4.44E-05 2.14E-05 9.07E-07 8.08E-11 0.00E+0 0.0OOE+00 6.67E-05 Te-132 0.OOE+00 0.00E+00 4.47E-04 2.21E-04 9.93E-06 1.11E-09 4.08E-1 1 0.00E+00 6.78E-04 1-131 2.72E-03 2.58E-03 2A.E-03 1.03E-03 1.58E-04 4.45E-05 1.22E-05 8.93E-07 9.02E-03 1-132 3.62E-03 3.26E-03 2.01E-03 4.59E-04 2.14E-05 5.15E-08 0.00E+00 0.00E+00 9.37E-03 1-133 5.65E-03 5.33E-03 4.90E-03 1.92E-03 2.61E-04 4.58E-05 IA17E-06 0.00E+00 1.81E-02 1-134 4.81E-03 3.99E-03 1.18E-03 1.01E-04 6.67E-07 0.00E+00 0.00E+00 0.00E+00 1.01E-02 1-135 5.20E-03 4.84E-03 4.00E-03 1.36E-03 1.38E-04 7.74E-06 1.44E-09 0.00E+00 1.55E-02 Xe-133 7.16E-03 1.03E-02 2.59E+00 4.89E+00 9.98E+00 6.96E+00 1.82E+00 4.04E-02 2.63E+01 Xe-135 1.66E-03 2.36E-03 5.35E-01 8.76E-01 1.35E+00 3.03E-01 4.86E-04 0.00E+00 3.06E+00 Cs-134 4.46E-04 4.23E-04 3.53E-04 1.32E-04 5.77E-06 7.39E-10 5.13E-1 1 3.24E-11 1.36E-03 Cs-136 1.20E-04 1.13E-04 9.43E-05 3.50E-05 1.52E-06 1.88E-10 1.12E-11 1.83E-12 3.64E-04 Cs-137 2.67E-04 2.53E-04 2.11 E-04 7.87E-05 3.45E-06 4.43E-10 3.08E-1 1 1.99E-11 8.13E-04 Ba-139 0.00E+00 0.OOE+00 8.89E-05 1.64E-05 1.02E-07 0.00E+00 0.00E+00 0.00E+00 1.05E-04 Ba- 140 0.OOE+00 0.00E+00 2.39E-04 1.20E-04 5.52E-06 6.87E-10 4.06E-1 1 6.40E-12 3.64E-04 La-140 0.00E+00 0.00E+00 2.36E-06 1. 15E-06 5.00E-08 0.00E+00 0.00E+00 0.00E+00 3.57E-06 La-141 0.00E+00 0.OOE+00 1.59E-06 5.63E-07 1.29E-08 0.00E+00 0.00E+00 0.00E+00 2.16E-06 La-142 0.00E+00 0.OOE+00 8.85E-07 1.82E-07 1.40E-09 0.00E+00 0.OOE+00 0.00E+00 1.07E-06 Ce-141 0.001E+00 0.OOE+00 5.43E-06 2.74E-06 1.27E-07 1.61E-1 1 1.05E-12 0.00E+00 8.30E-06 Ce-143 0.00E+00 0.00E+00 5.08E-06 2.46E-06 1.05E-07 0.OOE+00 0.00E+00 0.00E+00 7.65E-06 Ce-144 0.00E+00 0.00E+00 3.53E-06 1.78E-06 8.27E-08 1.07E-1 1 7.36E-13 4.48E-13 5.39E-06 Pr-143 0.00E+00 0.001E+00 2.07E-06 1.04E-06 4.79E-08 5.96E-12 0.00E+00 0.OOE+00 3.15E-06 Nd-147 0.00E+00 0.00E+00 9.23E-07 4.63E-07 2.13E-08 2.63E-12 0.00E+0 0.0OOE+00 1AIE-06 Np-239 0.001E+00 0.001E+00 6.74E-05 3.32E-05 1.47E-06 1.56E-10 0.00E+00 0.00E+00 1.02E-04 Pu-238 0.00E+00 0.00E+00 4.81E-09 2.42E-09 1.13E-10 1.45E-14 1.O1E-15 0.00E+00 7.34E-09 Pu-239 0.00E+00 0.00E+00 1.22E-09 6.15E-10 2.86E- 11 3.68E-15 2.56E-16 1.66E-16 1.86E-09 Pu-240 0.00E+00 0.001E+00 1.53E-09 7.69E-10 3.58E-1 1 4.61E-15 3.21E-16 2.08E-16 2.33E-09 Pu-241 0.00E+00 0.00E+00 2.63E-07 1.32E-07 6.16E-09 7.94E-13 5.52E-14 3.57E-14 4.01E-07 Am-241 0.00E+00 0.OOE+00 1.07E-10 5.39E-1 1 2.50E-12 3.23E-16 0.00E+0O 0.OOE+00 1.63E-10 Cm-242 0.OOE+00 0.00E+00 2.82E-08 1.42E-08 6.60E-10 0.00E+00 0.00E+00 O.OOE+00 4.3 1E-08 Cm-244 0.OOE+00 0.00E+00 1.52E-09 7.67E-10 3.57E-1 4.60E-15 0.00E+0O 0.OOE+00 2.32E-09 From RADTRAD Computer Run HASTI000CL03 I Nuclear Common Revision 9 1 Nula Cmo evso

CALULTION CONTINATION SHEET SHEET 61 of 84 CALC. NO.: H-I-ZZ-MDC-1880

REFERENCE:

LCR H02-01 & DCP 80048085 l ~G. Patel/NUCORE, ORIGINATOR, DATE REV: IIt27/02 M. Drucker/NUCORE, REVIEWERIVERIFIER, DATE 11/29/02 Table 12 Containment Leakage Total Activity on CR Charcoal/HEPA Filter (C0 Isotope 0-720 Isotope 0-720 Co-58 1.04E-07 Te-131m 4.89E-05 Co-60 1.25E-07 Te-132 4.98E-04 Kr-85 0.OOE+00 1-131 2.83E-03 Kr-85m O.OOE+00 1-132 1.863-03 Kr-87 0.00E+00 1-133 5AIE-03 Kr-88 0.00E+00 1-134 9.44E-04 Rb-86 1.24E-06 1-135 4.15E-03 Sr-89 1.51E-04 Xe-133 0.00E+00 Sr-90 1.07E-05 Xe-135 0.00E+00 Sr-91 1.61E-04 Cs-134 3.75E-04 Sr-92 1.03E-04 Cs-136 1.00E-04 Y-90 1.11E-07 Cs-137 2.24E-04 Y-91 1.85E-06 Ba-139 7.45EM05 Y-92 1.21E-06 Ba-140 2.68E-04 Y-93 1.94E-06 La- 140 2.62E-06 Zr-95 2.43E-06 La-141 1.56E-06 Zr-97 2.23E-06 La-142 7.40E-07 Nb-95 2.30E-06 Ce-141 6.10E-06 Mo-99 3.22E-05 Ce-143 5.61E-06 Tc-99m 2.08E-05 Ce-144 3.96E-06 Ru-103 2.52E-05 Pr-143 2.32E-06 Ru-105 1.09E-05 Nd-147 1.03E-06 Ru-106 6.85E-06 Np-239 7.49E-05 Rh-105 1.19E-05 Pu-238 5.40E-09 Sb-127 3.11E-05 Pu-239 1.37E-09 Sb-129 7.09E-05 Pu-240 1.711E-09 Te-127 2.50E-05 Pu-241 2.95E-07 Te-127m 4.14E-06 Am-241 1.20E-10 Te-129 2.26E-05 Cm-242 3.16E-08 Te-129m 2.71E-05 Cm-244 1.71E-09 I Nuclear Common Revision 9 I Nula omnRvso I

CALCULATION CONTINUATION SHEET SHEET 62 of 84 CALC. NO.: H-I-ZZ-MDC-1880

REFERENCE:

LCR H02-01 & DCP 80048085 G. Patel/NUCORE, ORIGINATOR, DATE REV: 11t27/02 I M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE I It29/02 Table 13 Post-LOCA MSIV Leakage Activity in CR With Charcoal/HEPA Filters (Ci)

Isotope 0-24 24-29.52 29.52-96 96-240 240-480 480-720 Total Kr-85 4.06E-02 1.37E-02 2.24E-02 1.65E-02 1.49E-02 1.35E-02 1.22E-01 Kr-85m 3.60E-02 5.18E-03 2.89E-07 0.00E+00 0.OOE+00 0.00E+00 4.12E-02 Kr-87 5.60E-06 9.33E-08 0.00E+00 0.00E+00 0.00E+00 0.00E+00 5.69E-06 Kr-88 1.04E-02 9.10E-04 0.00E+00 0.00E+00 0.00E+00 0.00E+00 1.13E-02 1-131 4.89E-03 9.50E-04 1.22E-03 3.15E-04 9.62E-05 2.26E-05 7.50E-03 1-132 5.66E-06 2.13E-07 0.00E+00 0.00E+00 0.00E+00 0.00E+00 5.87E-06 1-133 5.03E-03 8.29E-04 1.48E-04 5.26E-07 1.28E-10 0.OOE+00 6.01E-03 1-134 0.001E+00 0.00E+00 0.00E+00 0.0013E+00 O.1+00 0.O01+00 0.000E+

1-135 8.51E-04 9.45E-05 1.45E-07 0.00E+0 0.0OOE+00 0.00E+00 9.45E-04 Xe-133 7.71E+00 2.53E+00 2.86E+00 9.53E-01 2.31E-01 5.58E-02 1A3E+01 Xe-135 3.35E-01 7.43E-02 7.64E-04 9.57E-09 0.00E+00 0.OOE+00O 4.1OE-01 From RADTRAD Computer Run HAST1000MS02 I

I Nuclear Common Revision 9 1 I Nula I

omnRvso

CALCULATION CONTINUATION SHEET SHEET 63 of 84 CALC. NO.: H-I-ZZ-MDC-1880

REFERENCE:

LCR H02-01 & DCP 80048085 G. Patel/NUCORE, ORIGINATOR, DATE REV: 11/27/02 1 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 11/29/02 Table 14 Post-LOCA MSIV Leakage Activity in CR Without Charcoal/HEPA Filters Total Activity (Ci) C/HEPA Fltr Isotope 0-24 24-29.52 29.52-96 96-240 240-480 480-720 Total (Ci)

Kr-85 4.06E-02 1.37E-02 2.24E-02 1.65E-02 1.49E-02 1.35E-02 1.22E-01 0.00E+00 Kr-85m 3.60E-02 5.18E-03 2.89E-07 0.00E+00 0.00E+00 0.00E+00 4.12E-02 0.00E+00 Kr-87 5.60E-06 9.33E-08 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 5.69E-06 0.00E+00 Kr-88 1.04E-02 9.10E-04 0.00E+00 0.00E+00 0.00E+00 0.00E+00 1.13E-02 0.OOE+00 1-131 2.21E-02 4.31E-03 5.53E-03 1.43E-03 4.36E-04 1.02E-04 3.39E-02 2.64E-02 1-132 2.56E-05 9.64E-07 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 2.66E-05 2.07E-05 1-133 2.28E-02 3.76E-03 6.68E-04 2.38E-06 5.79E-10 5.79E-10 2.72E-02 2.12E-02 1-134 0.O0E1+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 0.00E+00 1-135 3.85E-03 4.28E-04 6.55E-07 0.00E+00 0.00E+00 0.00E+00 4.28E-03 3.33E-03 Xe-133 7.71E+00 2.53E+00 2.86E+00 9.53E-01 2.3 1E-01 5.58E-02 1.43E+01 0.00E+00 Xe-135 3.35E-01 7.43E-02 7.64E-04 9.57E-09 0.OOE+00 O.OOE+00 4.10E-01 0.00E+00 From RADTRAD Computer Run HASTI000MS03 I Nuclear Common Revision 9 1 I Nula!omnRvso

CALCULATION CONTINUATION SHEET SHEET 64 of 84 CALC. NO.: H-1-ZZ-MDC-1880

REFERENCE:

LCR H02-01 & DCP 80048085 G. Patel/NUCORE, ORIGINATOR, DATE REV: 11t27/02 1 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 11/29/02 Table 15 Post-LOCA ESF Leakage Activity In CR With Charcoal/HEPA Filters (Ci)

Isotope 0-0.33 0.33-0.5 0.5-2 2-4 4-8 8-24 24-96 96-720 Total 1-131 1.78E3-03 1.69E-03 5.43E-04 3.76E-04 3.41E-04 1.28E-04 7.01E-05 5.04E-06 4.93E-03 1-132 2.37E-03 2.13E-03 4.40E-04 1.68E-04 4.62E-05 1.49E-07 0.00E+00 0.00E+00 5.16E-03 1-133 3.70E-03 3.49E-03 1.07E-03 7.01E-04 5.63E-04 1.32E-04 8.47E-06 0.00E+00 9.66E-03 1-134 3.15E-03 2.61E-03 2.58E-04 3.71E-05 1.44E-06 0.OOE+00 0.00E+00 0.00E+00 6.06E-03 1-135 3.40E-03 3.17E-03 8.76E-04 4.95E-04 2.99E-04 2.23E-05 8.30E-09 O.OOE+00 18.26E-03 From RADTRAD Run HASTI000ESF02 Table 16 Post-LOCA ESF Leakage Activity in CR Without Charcoal/HEPA Filters Total Activity (Ci) C/HEPA Fltr Isotope 0-0.33 0.33-0.5 0.5-2 2-4 4-8 8-24 24-96 96-720 Total (CI) 1-131 1.78E-03 1.69E-03 1.95E-03 1.67E-03 1.54E-03 5.82E-04 3.17E-04 228E-05 9.55E-03 4.62E-03 1-132 2.37E-03 2.13E-03 1.58E-03 7.45E-04 2.09E-04 6.73E-07 0.00E+00 0.00E+00 7.03E-03 1.88E-03 1-133 3.70E-03 3.49E-03 3.85E-03 3.111E-03 2.55E-03 5.98E-04 3.84E-05 0.OOE+00 1.73E-02 7.67E-03 1-134 3.15E-03 2.61E-03 9.27E-04 1.65E-04 6.53E-0600.1+0 10.001E+00 10.003E+O0 6.86E-03 8.01E-04 1-135 3.40E-03 3.17E-03 3.14E-03 2.20E-03 1.36E-03 1.01E-04 3.76E-08 0.OOE+O0 1.34E-02 5.1OE-03 From RADTRAD Run HASTIOOOESF03 I Nuclear Common Revision 9 I I ula omnRvso

I CALCULATION CONTINUATION SHEET ISHEET 65 of 84 CALC. NO.: H-I-ZZ-MDC-1880

REFERENCE:

LCR H02-01 & DCP 80048085 G. Patel/NUCORE, ORIGINATOR, DATE REV: 11/27/02 1 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 11/29/02 Table 17 Post-LOCA Total Iodine & Aerosol Activity In CR Charcoal/HEPA Filters (Ci)

Containment MSIV ESF Total Leakage Leakage Leakage Iodine &

Isotope 0-720 0-720 0-720 Aerosol Co-58 1.04E-07 0.00E+00 0.001E3+0 1.04E-07 Co-60 1.25E-07 0.OOE+00 0.00E+00 1.25E-07 Kr-85 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Kr-85m 0.OOE+00 0.00E+00 O.OOE+00 0.00E+00 Kr-87 0.00E+00 0.00E+00 0.001E+00 0.00E+00 Kr-88 0.00E+00 0.00E+00 O.OOE+00 0.00E+00 Rb-86 1.24E-06 0.00E+00 O.OOE+00 1.24E-06 Sr-89 1.5 1E-04 0.00E+00 0.OOE+00 1.51E-04 Sr-90 1.07E-05 0.00E+00 O.OOE+00 1.07E-05 Sr-91 1.61E-04 0.00E+00 0.OOE+00 1.61E-04 Sr-92 1.03E-04 0.00E+00 0.OOE+00 1.03E-04 Y-90 1.11E-07 O.OOE+00 0.00E+00 1.I1E-07 Y-91 1.85E-06 0.00E+00 0.001E+00 1.85E-06 Y-92 1.21E-06 0.00E+00 0.00E+00 1.21E-06 Y-93 1.94E-06 0.OOE+00 0.00E+00 1.94E-06 Zr-95 2.431E-06 0.00E+00 0.00E+00 2.43E-06 Zr-97 2.23E-06 0.00E+00 0.003E+00 2.23E-06 Nb-95 2.30E-06 0.OOE+00 0.00E+00 2.30E-06 Mo-99 3.22E-05 0.00E+00 O.OOE+00 3.22E-05 Tc-99m 2.08E-05 0.OOE+00 0.001E+00 2.08E-05 Ru-103 2.52E-05 0.00E+00 0.00E+00 2.52E-05 Ru- 105 1.09E-05 0.00E+00 0.00E+00 1.09E-05 Ru- 106 6.851E-06 0.00E+00 0.OOE+00 6.85E-06 Rh- 105 1.19E-05 0.00E+00 0.00E+00 1.19E-05 Sb-127 3.11E-05 0.00E3+O0 0.00E+00 3.11 E-05 Sb-129 7.09E-05 0.00E+00 0.00E+00 7.09E-05 Te-127 2.50E-05 0.001E+00 0.00E+00 2.50E-05 Te-127m 4.14E-06 0.00E+00 0.00E+00 4.14E-06 Te- 129 2.26E-05 0.00E+00 0.00E+00 2.26E-05 Te-129m 2.71E-05 0.OOE+00 O.OOE+00 2.71E-05 I

I Nuclear Common Revision 9 i eiso Nula omo

CALCULATION CONTINUATION SHEET SHEET 66 of 84 CALC. NO.: H-I-ZZ-MDC-1880

REFERENCE:

LCR H02-01 & DCP 80048085 G. Patel/NUCORE, ORIGINATOR, DATE REV: 11/27/02 1 M. Drucker/NUCORE, REVIEWERNVERIFIER, DATE 11/29/02 Table 17 (Cont'd)

Post-LOCA Total Iodine & Aerosol Activity In CR Charcoal/HEPA Filters (Ci)

Containment MSIV ESF Total Leakage Leakage Leakage Iodine &

Isotope 0-720 0-720 0-720 Aerosol Te-131m 4.89E-05 0.00E+00 0.00E+00 4.89E-05 Te- 132 4.98E-04 0.00E+00 0.00E400 4.98E-04 1-131 2.83E-03 2.64E-02 4.62E-03 3.39E-02 1-132 1.86E-03 2.07E-05 1.88E-03 3.76E-03 1-133 5.41E-03 2.12E-02 7.67E-03 3.43E-02 1-134 9.44E-04 0.OOE+00 8.01E-04 1.75E-03 1-135 4.15E-03 3.33E-03 5.10E-03 11.6E-02 Xe-I 33 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 Xe-135 0.00E+00 O.OOE+00 0.00E+00 0.OOE+00 Cs-134 3.75E-04 0.OOE+00 0.00E+00 3.75E-04 Cs- 136 1.001E-04 0.00E+00 0.OOE+00 1.001E-04 Cs- 137 2.24E-04 0.00E+00 0.00E+00 2.24E-04 Ba-139 7.45E-05 0.00E+00 0.OOE+00 7.45E-05 Ba-140 2.68E-04 0.00E+00 0.00E+00 2.68E-04 La-140 2.62E-06 0.OOE+00 0.00E+00 2.62E-06 La- 141 1.56E-06 0.OOE+00 0.001E+00 1.56E-06 La-142 7.40E-07 0.00E+00 0.OOE+00 7.40E-07 Ce-141 6.10E-06 0.OOE+00 0.OOE+00 6.1OE-06 Ce-143 5.61E-06 0.00E+00 0.00E+00 5.61E-06 Ce-144 3.96E-06 0.00E+00 0.00E+00 3.96E-06 Pr-143 2.32E-06 0.00E+00 0.00E+00 2.32E-06 Nd-147 1.03E-06 0.00E+00 O.OOE+00 1.03E-06 Np-239 7.49E-05 0.00E+00 O.OOE+00 7.49E-05 Pu-238 5.40E-09 0.00E+00 0.00E+00 5.40E-09 Pu-239 1.37E-09 0.00E+00 0.OOE+00 1.37E-09 Pu-240 1.711E-09 0.003E+00 0.00E+00 1.711E-09 Pu-24 ! 2.95E-07 0.001E+00 0.00E+00 2.95E-07 Am-241 1.20E-10 0.00E+00 0.00E+00 1.20E-10 Cm-242 3.16E-08 0.00E+00 0.00E+00 3.16E-08 Cm-244 1.71E-09 0.00E+00 0.00E+00 1.7 IE-09 I Nuclear Common Revision 9 I NulalomnRvso

CALCULATION CONTINUATION SHEET SHEET 67 of 84 CALC. NO.: H-1-ZZ-MDC-1880

REFERENCE:

LCR H02-01 & DCP 80048085 G. Patel/NUCORE, ORIGINATOR, DATE REV: 11I27/02 1 M. Drucker/NUCORE, REVIEWERNVERIFIER, DATE 11)29/02 Table 18 0-1 Hr Isotopic Activity Released To Environment Isotopic Cumulative Integrated Activity Total Isotope Released To Environment (0-1 Hr) (Ci) Integrated Containment ESF MSIV Activity Leakage Leakage Leakage (Ci)

A B C A+B+C Co-58 8.397E-06 0.OOOE+00 0.OOOE+00 8.397E-06 Co-60 1.0053E-05 0.OOOE+00 0.OOOE+00 1.005E-05 Kr-85 2.673E+00 0.000E+00 0.000E+00 2.673E+00 Kr-85m 2.971E+01 0.0003E+00 O.OOOE+00 2.971E+01 Kr-87 4.102E+01 0.000E+3 0.0OOOE+00 4.102E+01 Kr-88 7.225E+01 O.O00E+00 0.O00E+O0 7.225E+01 Rb-86 6.150E-02 0.000E+00 0.000E+00 6.150E-02 Sr-89 9.032E-03 0.OOOE+00 0.000E+00 9.032E-03 Sr-90 1.666E-03 0.000E+00 0.000E+00 1.666E-03 Sr-91 1.740E-02 0.O00E+O0 O.OOOE+00 1.740E-02 Sr-92 1.040E-02 0.000E+00 0.000E+00 1.040E-02 Y-90 1.733E-05 0.000E+00 0.000E+00 1.733E-05 Y-91 1.208E-04 0.000E+00 0.000E+00 1.208E-04 Y-92 1. 104E-04 0.000E+00 0.000E+00 1.104E-04 Y-93 1.489E-04 0.000E+00 0.000E+00 1.489E-04 Zr-95 1.853E-04 0.O00E+O0 0.O00E+00 1.853E-04 Zr-97 5.533E-04 0.000E+00 0.000E+00 5.533E-04 Nb-95 1.861E-04 0.000E+00 0.000E+00 1.861E-04 Mo-99 2.872E-03 O.OOOE+00 0.000E+00 2.872E-03 Tc-99m 2.285E-03 0.000E+00 0.OOOE+00 2.285E-03 Ru-103 4.909E-03 0.OOOE+00 0.OOOE+00 4.909E-03 Ru-105 1.681E-03 0.OOOE+00 0.000E+00 1.681E-03 Ru-106 2.594E-03 0.OOOE+00 0.OOOE+00 2.594E-03 Rh-105 1.747E-03 0.000E+00 0.000E+00 1.747E-03 Sb-127 3.688E-03 0.O00E+00 O.OOOE+O0 3.688E-03 Sb-129 9.080E-03 0.0003E+00 0.000E+00 9.0801-03 Te-127 3.445E-03 0.000E+00 0.000E+00 3.445E-03 Te-127m 4.952E-04 0.000E+00 0.000E+00 4.952E-04 Te-129 6.007E-03 0.000E+00 0.000E+00 6.007E-03 Te-129m 1.538E-03 0.000E+00 0.000E+00 1.538E-03 I Nuclear Common Revision 9 1 I

I Nula!omnRvso

CALCULATION CONTINUATION SHEET SHEET 68 of 84 CALC. NO.: H-I-ZZ-MDC-1880

REFERENCE:

LCR H02-01 & DCP 80048085 G. Patel/NUCORE, ORIGINATOR, DATE REV: 11/27/02 1 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 11/29/02 Table 18 (Cont'd) 0-1 Hr Isotopic Activity Released To Environment Isotopic Cumulative Integrated Activity Total Isotope Released To Environment (0-1 Hr) (Ci) Integrated Containment ESF MSIV Activity Leakage Leakage Leakage (CO A B C A+B+C Te-131m 3.094E-02 0.000E+00 0.000E+00 3.094E-02 Te- 132 4.269E-02 0.000E+00 0.000E+00 4.269E-02 1-131 1.361E+01 1.428E9+00 0.OOOE+00 1.504E+01 1-132 1.871E+01 1.833E+00 0.000E+00 2.054E3+01 1-133 2.659E+01 2.770E+00 0.000E+00 2.936E+01 1-134 2.603E+01 2.321E+00 0.000E+00 2.835E+01 1-135 3.007E+01 3.080E+00 0.OOOE+00 3.315E+01 Xe-133 2.998E+02 0.000E+00 0.000E+00 2.998E+02 Xe-135 7.922E+01 0.000E+00 0.000E+00 7.922E+01 Cs-I134 6.2411E+00 0.OOOE+00 0.000E+00 6.241E+00 Cs-136 1.752E+00 0.000E+00 0.000E+00 1.752E3+00 Cs-137 5.186E+00 0.000E+00 0.000E+00 5.186E+00 Ba-139 1.322E-02 0.000E+00 0.000E+00 1.322E-02 Ba-140 2.013E-02 0.OOOE+00 0.000E+O0 2.013E-02 La-140 2.155E-04 0.000E+00 O.OOOE+00 2.155E-04 La-141 1.618E-04 0.000E+00 0.000E+00 1.618E-04 La-142 1.205E-04 0.000E+00 O.OOOE+00 1.205E-04 Ce-141 4.776E-04 0.000E+00 0.000E+00 4.776E-04 Ce-143 4.215E-04 0.000E+00 0.000E+00 4.215E-04 Ce-144 7.948E-04 0.000E+00 0.000E+00 7.948E-04 Pr-I143 1.660E-04 0.000E+00 0.OOOE+00 1.660E-04 Nd-147 7.818E-05 0.000E+00 0.OOOE+00 7.818E-05 Np-239 7.516E-03 0.000E+00 0.000E+00 7.516E-03 Pu-238 3.782E-06 0.OOOE+00 0.000E+00 3.782E-06 Pu-239 1.465E-07 0.000E+00 0.000E+00 1.465E-07 Pu-240 2.939E-07 0.000E+00 0.000E+00 2.939E-07 Pu-241 5.955E-05 0.000E+00 0.000E+00 5.955E-05 Am-241 3.194E-08 0.000E+00 0.000E+00 3.194E-08 Cm-242 1. 128E-05 0.000E+00 0.000E+00 1.128E-05 Cm-244 2.280E-06 0.000E+00 0.;'0E+0O 2.280E-06 I Nuclear Common Revision 9 1 I Nula!omnRvso

CALCULATION CONTINUATION SHEET SHEET 69 of 84 CALC. NO.: H-I-ZZ-MDC-1880

REFERENCE:

LCR H02-01 & DCP 80048085 G. PateM/UCORE, ORIGINATOR, DATE REV: 11/27/02 1 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 11/29/02 Table 19 0-2 Hr Isotopic Activity Released To Environment Isotopic Cumulative Integrated Activity Total Isotope Released To Environment (0-2 Hr) (Ci) Integrated Containment ESF MSIV Activity Leakage Leakage Leakage (Ci)

A B C A+B+C Co-58 1.517E-04 0.000E+00 0.000E+00 1.517E1-04 Co-60 1.8166E-04 0.000E+00 0.OOOE+00 1.816E-04 Kr-85 3.698E+01 0.000E+00 0.OOOE+00 3.698E+01 Kr-85m 3.615E+02 0.000E+00 0.000E+00 3.615E+02 Kr-87 3.633E+02 0.000E+00 0.OOOE+O0 3.633E+02 Kr-88 8.167E+02 0.000E+00 0.OOOE+00 8.167E+02 Rb-86 7.588E-02 0.000E+00 0.000E+00 7.588E-02 Sr-89 1.631E-01 0.000E+00 0.000E+00 1.631E-01 Sr-90 3.009E-02 0.000E+00 0.000E+00 3.009E-02 Sr-91 2.985E-01 0.000E+0 0.0OOOE+00 2.985E-01 Sr-92 1.572E-01 0.000E+0 0.0OOOE+00 1.572E-01 Y-90 3.106E-04 0.000E+O0 0.000E+00 3.106E-04 Y-91 2.182E-03 0.000E+00 0.000E+O0 2.182E-03 Y-92 1.739E-03 0.000E+00 0.000E+00 1.739E-03 Y-93 2.562E-03 0.000E+00 0.OOOE+00 2.562E-03 Zr-95 3.346E-03 0.000E+00 0.000E+00 3.346E-03 Zr-97 9.709E-03 0.O00E+O0 0.O00E+00 9.709E-03 Nb-95 3.360E-03 0.000E+00 0.000E+00 3.360E-03 Mo-99 5.150E-02 0.000E+00 0.000E+00 5.150E-02 Tc-99m 3.807E-02 0.000E+00 0.000E+00 3.807E-02 Ru-103 8.863E-02 0.000E+00 0.000E+00 8.863E-02 Ru- 105 2.72 IE-02 0.000E+00 0.000E+00 2.721E-02 Ru-106 4.686E-02 0.000E+00 0.000E+00 4.686E-02 Rh-105 3.112E-02 0.0003E+00 0.000E+00 3.112E-02 Sb-127 6.626E-02 0.000E+00 0.000E+00 6.626E-02 Sb- 129 1.466E-01 0.000E+00 0.000E+00 1A66E-01 Te-127 5.906E-02 0.000E+00 0.OOOE+00 5.906E-02 Te-127m 8.944E-03 0.000E+00 0.OOOE+00 8.944E-03 Te-129 7.221E-02 0.600E+00 0.000E+00 7.2211E-02 Te-129m 2.777E-02 0.000E+00 0.OOOE+00 2.777E-02 I Nuclear Common Revision 9 1 IN lerCm oReiin9I

CALCULATION CONTINUATION SHEET SHEET 70 of 84 CALC. NO.: H-I-ZZ-MDC-1880

REFERENCE:

LCR H02-01 & DCP 80048085 G. Patel/NUCORE, ORIGINATOR, DATE REV: 11/27/02 I M. Drucker/NUCORE, REVIEWERIVERIFIER, DATE 11/29/02 Table 19 (Cont'd) 0-2 Hr Isotopic Activity Released To Environment Isotopic Cumulative Integrated Activity Total Isotope Released To Environment (0-2 Hr) (Ci) Integrated Containment ESF MSIV Activity Leakage Leakage Leakage (Ci)

A B C A+B+C Te-13 Im 5.499E-01 0.000E+00 0.OOOE+00 5.499E-01 Te-132 7.664E-01 0.000E+00 0.OOOE+00 7.664E-01 1-131 2.073E+01 6.283E+00 0.OOOE+00 2.702E+01 1-132 2.499E+01 6.097E+00 0.000E+00 3.108E+01 1-133 3.990E+01 1.184E+01 0.000E+00 5.174E+01 1-134 3.029E+01 5.190E+00 0.000E+00 3.548E+01 1-135 4.360E+01 1.229E+01 0.000E+00 5.589E+01 Xe-133 4.128E+03 0.OOOE+00 0.000E+00 4.128E+03 Xe- 135 1.029E+03 0.000E+00 0.000E+O0 1.029E+03 Cs-134 7.704E+00 0.0OOE+00 0.000E+00 7.704E+00 Cs- 136 2.162E+00 0.0001E+00 0.000E+00 2.162E+00 Cs-137 6.402E+00 0.000E+00 0.000E+00 6.402E+00 Ba- 139 1.691E-01 0.000E+00 0.000E+00 1.691E-01 Ba-140 3.63 IE-01 0.000E+00 0.OOOE+00 3.6311E-01 La-140 3.846E-03 0.000E+00 0.000E+00 3.846E-03 La-141 2.583E-03 0.000E+00 0.OOOE+00 2.583E-03 La-142 1.598E-03 0.000E+00 0.000E+00 1.598E-03 Ce-141 8.622E-03 0.000E+00 0.000E+00 8.622E-03 Ce-143 7.501E-03 0.000E+00 0.000E+00 7.501E-03 Ce-144 1.436E-02 0.000E+00 0.000E+00 1A36E-02 Pr-143 2.993E-03 0.000E+00 0.000E+00 2.993E-03 Nd-147 1.410E-03 0.000E+00 0.OOOE+00 1.410E-03 Np-239 1.346E-01 0.000E+00 0.0003E+00 1.346F3-01 Pu-238 6.832E-05 0.000E+00 0.000E+00 6.832E-05 Pu-239 2.646E-06 0.000E+00 0.000E+00 2.646E-06 Pu-240 5.3 1OE-06 0.0001E+00 0.000E+00 5.3103E-06 Pu-241 1.076E-03 0.000E+00 0.000E+00 1.076E-03 Am-241 5.769E-07 0.000E+00 0.000E+00 5.769E-07 Cm-242 2.038E-04 0.OOOE+00 0.000E+00 2.038E-04 Cm-244 4.119E-05 0.000E+00 0.OOOE+00 4.119E-05 I Nuclear Common Revision 9 1 Nula omn eiin9I

Table 20 0-4 Hr Isotopic Activity Released To Environment Isotopic Cumulative Integrated Activity Total Isotope Released To Environment (0-4 Hr) (Ci) Integrated Containment ESF MSIV Activity Leakage Leakage Leakage (Ci)

A B C A+B+C Co-58 3.870E-04 0.000E+00 0.000E+00 3.870E-04 Co-60 4.636E-04 0.000E+00 0.000E+00 4.636E-04 Kr-85 1.952E+02 0.000E+00 0.000E+00 1.952E+02 Kr-85m 1.576E+03 0.000E+00 0.000E+00 1.576E+03 Kr-87 1.040E+03 0.000E+0O 0.OOOE+00 1.040E+03 Kr-88 3.207E+03 0.000E+00 0.000E+00 3.207E+03 Rb-86 9.532E-02 0.000E+00 0.000E+00 9.532E-02 Sr-89 4.161E-01 0.000E+00 0.000E+00 4.161E-01 Sr-90 7.681E-02 0.000E+00 0.000E+00 7.681E-02 Sr-91 7.224E-01 0.000E+00 0.000E+00 7.224E-01 Sr-92 3.367E-01 0.000E100 0.000E+00 3.367E-01 Y-90 7.864E-04 0.000E+00 0.000E+00 7.864E-04 Y-91 5.567E-03 0.000E+00 0.0001E+00 5.567E-03 Y-92 3.871E-03 0.000E+00 0.000E+00 3.871E-03 Y-93 6.22 1E-03 0.000E+00 0.000E+00 6.2211E-03 Zr-95 8.537E-03 0.000E+00 0.000E+00 8.537E-03 Zr-97 2.405E-02 0.000E+00 0.000E+00 2.405E-02 Nb-95 8.570E-03 0.000E+00 0.0001E+00 8.570E-03 Mo-99 1.304E-01 0.OOOE+00 0.000E+00 1.304E-01 Tc-99m 8.945E-02 0.000E+00 0.0003E+00 8.945E-02 Ru-103 2.261E-01 0.OOOE+00 0.000E+00 2.261E-01 Ru-105 6.219E-02 0.000E+00 0.000E+00 6.219E-02 Ru-106 1.196E-01 0.000E+00 0.000E+00 1.196E-01 Rh-105 7.829E-02 0.000E+00 O.OOOE+00 7.829E-02 Sb-127 1.682E-101 0.000E+00 0.0003E+00 1.682E-01 Sb-129 3.339E-01 0.000E+00 0.000E+00 3.339E-01 Te-127 1.428E-01 0.0001E00 O.OOOE+O0 1.428E-01 Te-127m 2.283E-02 0.000E+00 0.000E+00 2.283E-02 Te-129 1.280E-01 0.000E+00 0.0001E+00 1.280E-01 Te-129m 7.084E-02 0.00013E+ 0.000E+0O 7.084E-02 I Nuclear Common Revision 9 eiso I Nula omo

Table 20 (Cont'd) 0-4 Hr Isotopic Activity Released To Environment Isotopic Cumulative Integrated Activity Total Isotope Released To Environment (0-4 Hr) (Ci) Integrated Containment ESF MSIV Activity Leakage Leakage Leakage (Ci)

A B C A+B+C Te-131m 1.380E+00 0.000E+00 0.000E+00 1.380E+00 Te-132 1.943E+00 0.000E+00 0.000E+00 1.943E+00 1-131 4.001E+01 2.527E+01 0.000E+00 6.529E+01 1-132 3.615E+01 1.681E+01 0.000E+00 5.296E+01 1-133 7.441E+01 4.574E+01 0.000E+00 1.202E+02 1-134 3.427E+01 8.851E+00 0.000E+00 4.312E+01 1-135 7.525E+01 4.319E+01 0.OOOE+00 1.184E+02 Xe-133 2.164E+04 0.000E+00 0.000E+00 2.164E+04 Xe-135 4.933E+03 0.000E+00 0.000E+00 4.933E+03 Cs-134 9.685E+00 0.000E+00 0.000E+00 9.685E+00 Cs-136 2.714E+00 0.000E+00 0.000E+00 2.714E+00 Cs-137 8.047E+00 0.000E+00 0.000E+00 8.047E+00 Ba-139 3.144E-01 0.000E+00 0.000E+00 3.144E-01 Ba-140 9.252E-01 0.000E+00 0.000E+00 9.252E-01 La-140 9.692E-03 0.OOOE+00 0.000E+00 9.692E-03 La-141 5.822E-03 0.000E+00 0.000E+00 5.822E-03 La-142 3.055E-03 0.000E+00 0.000E+00 3.055E-03 Ce-141 2.199E-02 0.000E+00 0.000E+00 2.199E-02 Ce-143 1.885E-02 0.000E+00 0.000E+00 1.885E-02 Ce-144 3.664E-02 0.0001+00 0.000E+00 3.664E-02 Pr-143 7.628E-03 0.000E+00 0.000E+00 7.628E-03 Nd-147 3.591E-03 0.000E+00 0.OOOE+00 3.591E-03 Np-239 3.40413-01 0.000E+00 O.OOOE+O0 3.404E-01 Pu-238 1.744E-04 0.000E+00 0.000E+00 1.744E-04 Pu-239 6.753E-06 0.000E+00 0.000E+00 6.753E-06 Pu-240 1.355E-05 0.000E+00 0.000E+00 1.355E-05 Pu-241 2.745E-03 0.000E+00 0.000E+00 2.745E-03 Am-241 1.473E-06 0.000E+00 0.000E+00 1.473E-06 Cm-242 5.200E-04 0.000E+00 0.000E+00 5.200E-04 Cm-244 1.05 IE-04 0.000E+00 0.000E+00 1.05 1E-04 I Nuclear Common Revision 9 1 I Nula omnRvso

CALCULATION CONTINUATION SHEET SHEET 73 of 84 CALC. NO.: H-1-ZZ-MDC-1880

REFERENCE:

LCR H02-01 & DCP 80048085 G. Patel/NUCORE, ORIGINATOR, DATE REV: 11/27/02 1 M. Drucker/NUCORE, REVIEWERNVERIFIER, DATE 11/29/02 Table 21 0-8 Hr Isotopic Activity Released To Environment Isotopic Cumulative Integrated Activity Total Isotope Released To Environment (0-8 Hr) (Ci) Integrated Containment ESF MSIV Activity Leakage Leakage Leakage (CO)

A B C A+B+C Co-58 4.377E1-04 0.000E+00 0.000E+00 4.377E-04 Co-60 5.245E-04 0.000E+00 0.OOOE+00 5.245E-04 Kr-85 8.297E+02 0.000E+00 0.000E+00 8.297E+02 Kr-85m 4.641E+03 0.000E+00 0.000E+00 4.641E+03 Kr-87 1.628E+03 0.OOOE+00 0.OOOE+00 1.628E+03 Kr-88 7.875E+03 0.000E+00 0.OOOE+00 7.875E+03 Rb-86 9.940E-02 0.OOOE+00 0.000E+00 9.940E-02 Sr-89 4.707E-01 0.000E+00 0.000E+00 4.707E-01 Sr-90 8.690E-02 0.OOOE+00 0.000E+00 8.690E-02 Sr-91 8.004E-01 0.000E+00 0.000E+00 8.004E-01 Sr-92 3.591E-01 0.OOOE+00 0.0003E+00 3.591E-01 Y-90 8.867E-04 0.0003E+00 0.000E+00 8.867E-04 Y-91 6.297E-03 0.000E+00 0.000E+00 6.297E-03 Y-92 4.173E-03 0.0003E+00 0.000E+00 4.173E-03 Y-93 6.901E-03 0.000E+00 0.000E+00 6.901E-03 Zr-95 9.657E-03 0.000E+00 0.000E+00 9.657E-03 Zr-97 2.688E-02 0.000E+00 0.000E+00 2.688E-02 Nb-95 9.693E-03 0.000E+00 0.000E+00 9.693E-03 Mo-99 1.4711E-01 0.OOOE+00 0.000E+00 1.471E-01 Tc-99m 9.808E-02 0.000E+00 0.000E+00 9.808E-02 Ru-103 2.557E-01 0.000E+00 0.000E+00 2.557E-01 Ru-105 6.757E-02 0.000E+00 0.OOOE+00 6.757E-02 Ru-106 1.353E-01 0.OOOE+00 0.OOOE+00 1.353E-01 Rh-105 8.804E-02 0.000E+00 0.000E+00 8.804E-02 Sb-127 1.898E-01 0.000E+00 0.000E+00 1.898E-01 Sb-129 3.625E-01 0.000E+00 0.000E+00 3.625E-01 Te-127 1.582E-01 0.000E+00 0.000E+00 1.582E-01 Te-127m 2.582E-02 0.000E+0 0.0OOOE+00 2.582E-02 Te-129 1.314E-01 0.000E+00 0.000E+00 1.314E-01 Te-129m 8.012E-02 0.000E+00 0.000E+00 8.012E-02 I Nuclear Common Revision 9 1 eiso Nula omo

CALCULATION CONTINUATION SHEET SHEET 74 of 84 CALC. NO.: H-I-ZZ-MDC-1880

REFERENCE:

LCR H02-01 & DCP 80048085 G. Patel/NUCORE, ORIGINATOR, DATE REV: 11/27/02 1 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 11/29/02 Table 21 (Cont'd) 0-8 Hr Isotopic Activity Released To Environment Isotopic Cumulative Integrated Activity Total Isotope Released To Environment (0-8 Hr) (Ci) Integrated Containment ESF MSIV Activity Leakage Leakage Leakage (Ci)

A B C A+B+C Te-131m 1.550E+00 O.OOOE+00 0.000E+00 1.550E+00 Te-132 2.193E+00 O.OOOE+00 0.OOOE+00 2.193E+00 1-131 9.671E+01 9.799E+01 0.000E+00 1.947E1+02 1-132 4.963E+01 3.398E+01 O.OOOE+00 8.361E+01 1-133 1.669E+02 1.643E+02 0.000E+00 3.312E+02 1-134 3.560E+01 1.052E+01 O.OOOE+00 4.613E+01 1-135 1.434E+02 1.304E+02 0.000E+00 2.737E+02 Xe-133 9.068E+04 0.000E+00 O.OOOE+00 9.068E+04 Xe-135 1.736E+04 0.000E+00 0.000E+00 1.736E+04 Cs-134 1.010E+01 O.OOOE+00 O.OOOE+00 1.0101E+01 Cs-136 2.830E+00 0.000E+00 O.OOOE+00 2.830E+00 Cs-137 8.394E+00 O.OOOE+00 0.OOOE+00 8.394E+00 Ba-139 3.252E-01 O.OOOE+00 O.OOOE+00 3.252E-01 Ba-140 1.046E+00 0.000E+00 0.000E+00 1.046E+00 La-140 1.091E&02 0.000E+00 O.OOOE+00 1.091E-02 La-141 6.300E-03 O.OOOE+00 O.OOOE+00 6.300E-03 La-142 3.177E-03 0.000E+00 0.000E+00 3.177E-03 Ce-141 2.487E-02 0.000E+00 0.000E+00 2.487E-02 Ce-143 2.119E-02 0.000E+00 0.000E+00 2.119E-02 Ce-144 4.145E-02 0.000E+00 0.OOOE+00 4.145E-02 Pr-143 8.624E-03 O.OOOE+00 0.OOOE+00 8.624E-03 Nd-147 4.059E-03 0.000E+00 O.OOOE+00 4.059E-03 Np-239 3.837E-01 O.OOOE+00 0.000E+00 3.837E-01 Pu-238 1.973E-04 O.OOOE+00 0.000E+00 1.973E-04 Pu-239 7.641 E-06 0.000E+00 0.000E+00 7.6411E-06 Pu-240 1.533E-05 0.000E+00 0.000E+00 1.533E-05 Pu-241 3.106E-03 0.000E+00 0.000E+00 3.106E-03 Am-241 1.666E-06 0.000E+00 O.OOOE+00 1.666E-06 Cm-242 5.S83E-04 0.000E+00 0.OOOE+00 5.883E-04 Cm-244 1.190E-04 0.000E+00 0.0O0E+00 1.190E-04 Revision 9 I I Nuclear Common Nuclear Common Revision 9 1

CALCULATION CONTINUATION SHEET SHEET 75 of 84 CALC. NO.: H-1-ZZ-MDC-1880

REFERENCE:

LCR H02-01 & DCP 80048085 G. Patel/NUCORE, ORIGINATOR, DATE REV: 11/27/02 1 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 11/29/02 Table 22 0-720 Hr Isotopic Activity Released To Environment Isotopic Cumulative Integrated Activity Total Isotope Released To Environment (0-720 Hr) (Ci) Integrated Containment ESF MSIV Activity Leakage Leakage Leakage (CQ)

A B C A+B+C Co-58 4.408E-04 0.000E+00 2.395E-02 2A39E-02 Co-60 5.281E-04 0.000E+00 2.904E-02 2.957E-02 Kr-85 1.392E+05 0.000E+00 4.749E+05 6.141E+05 Kr-85m 1.081E+04 0.000E+00 IA92E+04 2.573E+04 Kr-87 1.751E+03 0.000E+00 2.065E+02 1.957E+03 Kr-88 1.259E+04 0.000E+00 1.079E+04 2.338E+04 Rb-86 9.963E-02 0.000E+00 2.052E+00 2.151E+00 Sr-89 4.739E-01 0.000E+00 2.563E+01 2.61 IE+0I Sr-90 8.750E-02 0.000E+00 4.814E+00 4.901E+00 Sr-91 8.036E-01 0.0OOE+00 1.297E+01 1.378E+01 Sr-92 3.595E-01 0.0OOE+00 8.122E-01 1.172E+00 Y-90 8.922E-04 0.000E+00 3.745E-02 3.834E-02 Y-91 6.340E-03 0.000E+00 3.437E-01 3.501E-01 Y-92 4.180E-03 0.000E+00 1.714E-02 2.132E-02 Y-93 6.929E-03 0.000E+00 1.183E-01 1.252E-01 Zr-95 9.724E-03 0.000E+00 5.278E-01 5.375E-01 Zr-97 2.701E-02 0.000E+00 6.735E-01 7.005E-01 Nb-95 9.760E-03 0.000E+00 5.240E-01 5.337E-01 Mo-99 1.480E-01 0.0OOE+00 6.256E+00 6.404E+00 Tc-99m 9.837E-02 0.000E+00 9.584E-01 1.05713+00 Ru-103 2.575E-01 0.000E+00 1.386E+01 1.412E+01 Ru-105 6.772E-02 0.000E+00 4.202E-01 4.879E-01 Ru-106 1.362E-01 0.000E+00 7A78E+00 7.615E+00 Rh-105 8.856E-02 0.0003E+00 3.129E+00 3.217E+00 Sb-127 1.911 E-01 0.000E+00 8.633E+00 8.824E+00 Sb-129 3.633E-01 0.000E+00 2.152E+00 2.516E+00 Te-127 1.588E-01 0.000E+00 2.526E+00 2.685E+00 Te-127m 2.600E-02 0.000E+00 1.4199E+00 1.445E+00 Te-129 1.314E-01 0.000E+00 1.358E-02 1.450E-01 Te-129m 8.067E-02 0.000E+00 4.326E1+00 4.407E+00 I Nuclear Common Revision 9 1 evso Nula Cmo

CALCULATION CONTINUATION SHEET ISHEET 76 of 84 CALC. NO.: H-I-ZZ-MDC-1880

REFERENCE:

LCR H02-01 & DCP 80048085 G. Patel/NUCORE, ORIGINATOR, DATE REV: 11/27/02 1 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 11/29/02 Table 22 (Cont'd) 0-720 Hr Isotopic Activity Released To Environment Isotopic Cumulative Integrated Activity Total Isotope Released To Environment (0-720 Hr) (Ci) Integrated Containment ESF MSIV Activity Leakage Leakage Leakage (Ci)

A B C A+B+C Te-131m 1.559E+00 0.000E+00 5.174E+01 5.330E+01 Te-132 2.207E+00 0.000E+00 9.667E+01 9.888E+01 1-131 4.597E+03 1.122E+04 2.470E+04 4.051E+04 1-132 5.881E+01 4.598E+01 8.619E+01 1.910E+02 1-133 1.185E+03 1.944E+03 8.689E+03 1.182E+04 1-134 3.569E+01 1.065E+01 5.184E-01 4.686E+01 1-135 3.649E+02 4.382E+02 2.152E+03 2.955E+03 Xe-133 4.100E+06 0.000E+00 1.350E+07 1.760E+07 Xe-135 7.8101E+04 0.000E+00 1.532E+05 2.313E+05 Cs-134 1.013E+01 0.0003E+00 2.185E+02 2.286E+02 Cs-136 2.837E+00 0.000E+00 5.729E+01 6.012E+01 Cs-137 8.415E+00 0.000E+00 1.818E+02 1.902E+02 Ba-139 3.253E-01 0.OOOE+00 7.656E-02 4.0181E-01 Ba-140 1.053E+00 0.000E+00 5.427E+01 5.532E+01 La-140 1.097E-02 0.000E+00 4.053E-01 4.162E-01 La-141 6.312E-03 0.000E+00 3.163E-02 3.794E-02 La-142 3.177E-03 0.000E+00 1.198E-03 4.375E-03 Ce-141 2.505E-02 0.000E+00 1.342E+00 1.367E+00 Ce-143 2.131E-02 0.000E+00 7.341E-01 7.554E-01 Ce-144 4.174E-02 0.000E+00 2.290E+00 2.331E+00 Pr-143 8.682E-03 0.000E+00 4.491E-01 4.578E-01 Nd-147 4.087E-03 0.000E+00 2.085E-01 2.126E-01 Np-239 3.861E-01 0.000E+00 1.572E+01 1.6111E+01 Pu-238 1.987E-04 0.000E+00 1.093E-02 1.1 13E-02 Pu-239 7.694E-06 0.000E+00 4.233E-04 4.3 1OE-04 Pu-240 1.544E-05 0.000E+00 8.495E-04 8.649E-04 Pu-241 3.128E-03 0.000E+00 1.7211E-01 1.752E-01 Am-241 1.678E-06 0.0003E+00 9.230E-05 9.397E-05 Cm-242 5.924E-04 0.000E+00 3.242E-02 3.301E-02 Cm-244 1. 198E-04 0.0003E+00 6.589E-03 6.709E-03 I Nuclear Common Revision 9 1 eiin9I Nula omn

12.0 FIGURES Figure 1: Containment Leakage RADTRAD Nodalization Revision 9 I II Nuclear Common Revision 9 1

CALCULATION CONTINUATION SHEET SHEET 78 of 84 CALC. NO.: H-I-ZZ-MDC-1880

REFERENCE:

LCR H02-01 & DCP 80048085 G. Patel/NUCORE, ORIGINATOR, DATE REV: 11/27/02 1 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 11/29/02 Figure 2: HCGS ESF Leakage RADTRAD Nodalization I Nuclear Common Revision 9 1 I

I Nula!omnRvso

CALCULATION CONTINUATION SHEET SHEET 79 of 84 CALC. NO.: H-I-ZZ-MDC-1880

REFERENCE:

LCR H02-01 & DCP 80048085 G. Patel/NUCORE, ORIGINATOR, DATE REV: 11/27/02 1 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 11/29/02 Fieure 3: HCGS MSIV Leak RADTRAD Nodalization I Nuclear Common Revision 9 1 I Nula!omnRvso

CALCULATION CONTINUATION SHEET SHEET 80 of 84 CALC. NO.: H-I-ZZ-MDC-1880

REFERENCE:

LCR H02-01 & DCP 80048085 G. Patel/NUCORE, ORIGINATOR, DATE REV: 11/27/02 1 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 11/29/02 Figure 4 - HCGS Control Room RADTRAD Nodalization Revision 9 I II Nuclear Common Revision 9 1

CALCULATION CONTINUATION SHEET SHEET 81 of 84 CALC. NO.: H-I-ZZ-MDC-1880

REFERENCE:

LCR H02-01 & DCP 80048085 G. Patel/NUCORE, ORIGINATOR, DATE REV: 11/27/02 1 M. Drucker/NUCORE, REVIEWERNERIFIER, DATE 11/29/02 EL 158'-9" 10 [Y]

EL 155- -9""

EL 155--3" EL 154'-3" Rt CR Receptor Location (EL 143'-O")

Figure 5 - CR Filter Shine Dose (Elevation View)

I Nuclear Common Revision 9 I Nula omnRvso

CALCULATION CONTINUATION SHEET SHEET 82 of 84 CALC. NO.: H-I-ZZ-MDC-1880

REFERENCE:

LCR H02-01 & DCP 80048085 G. Patel/NUCORE, ORIGINATOR, DATE REV: 11/27/02 1 M. Drucker/NUCORE, REVIEWERNERIFIER, DATE 11/29/02 60.69 cms 121.92 cms 4 45.72

• cms

......... -91.44 10 ems X Indicates Dose Point Location Figure 6 - CR Filter Shine Dose (Plan View)

I Nuclear Common Revision 9 1 I Nula omnRvso

13.0 AFFECTED DOCUMENTS:

Upon approval of Licensing Change Request LCR H02-01, the following documents will be revised:

1. HCGS UFSAR Section 15.6.5

2. UFSAR Table 15.6-12

3. UFSAR Table 15.6-13

4. UFSAR Table 15.6-14

5. UFSAR Table 15.6-15

6. UFSAR Table 15.6-16 Upon approval of Licensing Change Request LCR H02-01, the following documents will be deleted:

1. UFSAR Table 15.6-17

2. UFSAR Table 15.6-18

3. UFSAR Table 15.6-19

4. UFSAR Table 15.6-20 14.0 ATTACHMENTS 14.1 E-mail

Subject:

FRVS Vent Charcoal Filter Efficiencies 14.2 E-mail

Subject:

MSIV Type C Leak Rate Test 14.3 - 2 Diskettes with the following electronic files:

Calculation No: H-1-ZZ-MDC-1880, Rev IlRi.

Comment Resolution Form 2 - Mark Drucker (llR2}

Comment Resolution Form 2 - John F. Duffy (OIR1 through lIR1)

HCGSMHADEF.txt For Cont., ESF, & MSIV Leakages Cont. Leakage RADTRAD Input File HEPU350CLOO.PSF Cont. Leakage RADTRAD Output File HEPU350CLOO.oO ESF Leakage RADTRAD Input File HEPU350ESFOO.PSF ESF Leakage RADTRAD Output File HEPU350ESFOO.oO MSIV Leakage RADTRAD Input File HEPU350MSOO.PSF MSIV Leakage RADTRAD Output File HEPU35OMSOO.oO Cont. Leakage RADTRAD Input File (HAST1000CL02.PSF/ HAST1000CL03.PSF)

Cont. Leakage RADTRAD Output File (HAST1000CL02.O0/HAST1000CLO3.O0)

ESF Leakage RADTRAD Input File (HAST1000ESF02.PSF/ HAST1000ESF03.PSF)

ESF Leakage RADTRAD Output File (HAST1000ESF02.O0/HAST1000ESF03.O0)

MSIV Leakage RADTRAD Input File (HAST1000MS02.PSF/ HAST1000MS03.PSF)

MSIV Leakage RADTRAD Output File (HAST1000MS02.O0[HAST1000MS03.O0)

MicroShield Input/Output Files (Hcgs.MS5)

Nuclear Common Revision 9

CALCULATION CONTINUATION SHEET SHEET 84 of 84 CALC. NO.: H-I-ZZ-MDC-1880

REFERENCE:

LCR H02-01 & DCP 80048085 G. Patel/NUCORE, ORIGINATOR, DATE REV: 11127/02 1 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE I 1/29/02 RADTRAD/HABIT1.0 V&V Files 990/ooa.DSG 99%/xxx.DSG HCCLI30MS00.PSF 99'Yoacb.INP 99%/xxxcb.1NP HCGSTIDDEF 99%/oacb.SPD 99%/xxxcb.SPD HCI30ESFOO.PSF 990/ooat5a.INP 99%/xxxt5a.TAB HCI30ESF00.O0 99%/oat5a.NUC 99%/xxx5a.INP HCCLI30MSOO.O0 990/ooat5a.TAB 99%/xxxt5a.NUC eiso omo I Nula Nuclear Common Revision 9 1