ML102371018

Calculation H-1-ZZ-MDC-1880, Rev. 4, Post-LOCA Eab, LPZ & CR Doses

ML102371018
Person / Time
Site:	Hope Creek
Issue date:	08/10/2010
From:	Gita Patel Public Service Enterprise Group
To:	Office of Nuclear Reactor Regulation
References
H-1-ZZ-MDC-1880, Rev 4
Download: ML102371018 (108)

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{{#Wiki_filter:Attachment 2 LR-NI10-0306 Calculation H-1-ZZ-MDC-1880, Revision 4, "Post-LOCA EAB, LPZ, and CR Doses"

CC-AA-309-1001-F1 Revision 0 Design Analysis Major Revision Cover Sheet Design Analysis (Major Revision) I Last Page No. 117 Analysis No.:' H-1-ZZ-MDC-1880 Revision: 2 4

Title:

3 Post-LOCA EAB, LPZ, and CR Doses DCP No(s),/ Revision: 4 DCP 80102144 Revision 0 AD No(s)./ Revision: s M01 Revision 0 Station(s): 7 Hope Creek Component(s): 114 Unit No.: 8 1 N/A Discipline:; Mechanical Design .. . . ..... ..... SafetylQA Class: 'j SR System Code: 12 NIA Structure: 13 N/A CONTROLLED DESIGN INPUTS AND OUTPUTS$6 Document No,: Fromlro Document No.: From/To H.I-ZZ-MDC-1879, Rev I From HAI.-ZZ-MDC.1923 To GE-NE-T2300789-00-02 From H-I-ZZ-MDC-I927 To H-I-ZZ.MDC-0384, Rev 1 From VHC-MD-ST.GK-0002 To For remaining References see Section 10 From Is this Design Analysis Safeguards Information? Yes El No [ If yes, see SY-AA-101-106 17 Does this Design Analysis contain Unverified Assumptions? Yes E No [ If yes, Order#: This Design Analysis SUPERCEDES: "8 H-1 -ZZ-MDC-1 880, Rev 3 inIts entirety. Description of Revision (list affected pages for partials): Is Revision 4 deletes the evaluation of the radiological Impact of keeping the PCIVs open for 120 seconds following a design basis LOCA, revises the MSIV model to include elements of the Revision 2 model that was reviewed by the NRC staff as part of the EPU License Amendment 174 (ADAMS.ML081230640), and reduces the control room envelope unfiltered in-leakage rate to 300 cfm. Preparer: 21 Gopal J. Patel (NUCORE) 0811012010 Print N e . . $I" Name ZI DAth Method of Review: 21 Detailed Review 0 Alternate Calculations (ttacýhed) 0 Testing 0 Reviewer: 22 Mark Drucker (NUCORE) 08/1012010 Print Name Sg Name nate Review Notes: Independent review [0 Peer review E] Mark Drucker WNUCORE _ 08110.2010 Print Name Sign Na e Date t(Ft etiemniAnalysan Only)... ..... External Approver: 2' NIA Print Name Sign Name Date PSEG Reviewer: 21 John Duffy I ViJay Chandra .'-,, ., w ',,._ 8/1012010 Print Nows, Namen V P at Independent 3rd Party Review Reqd? 2 Yes El No 0 PSEG Approver: James Boyer 811012010 Print Name .. .. ,', Date i 9

ATTACHMENT 2 Owners Acceptance Review Checldist for External Design Analysis Page 1 of 1 DESIGN ANALYSIS NO. H-1-ZZ-MDC-1880 REV: 4 Yes No N/A

1. Do assumptions have sufficient rationale? FD Are assumptions compatible with the way the plant is operated and with the licensing basis? E- Dl

3. Do the design inputs have sufficient rationale? D

4. Are design inputs correct and reasonable?

Are design inputs compatible with the way the plant is operated and with the licensing basis? Z 0 D Are Engineering Judgments clearly documented and justified? D E DE Are Engineering Judgments compatible with the way the plant is operated and with the licensing basis? 0l D Do the results and conclusions satisfy the purpose and objective of the Design Analysis? "l F-1 L Are the results and conclusions compatible with the way the plant is operated and with the licensing basis? " El El Does the Design Analysis include the applicable design basis documentation? Have any limitations on the use of the results been

11. identified and transmitted to the appropriate El r_1 organizations?

12. Are there any unverified assumptions?

Do all unverified assumptions have a tracking and closure mechanism in place? El El Z Have all affected design analyses been documented on the

14. Affected Documents List (ADL) for the associated El 0 r Configuration Change?

Do the sources of inputs and analysis methodology used meet current technical requirements and regulatory

15. commitments? (If the input sources or analysis methodology are based on an out-of-date methodology or Z El El code, additional reconciliation may be required if the site has since committed to a more recent code)

Have vendor supporting technical documents and references (including GE DRFs) been reviewed when necessary? El El

17. Has the Vendor supplied the native electronic file(s)? El El PSEG REVIEWER: John. Duffy*/f. %1 DATE: 8/10 /2010 Print / Sign M. Drucker of NUCORE performed a detailed review. Additionally, A. Klazura of S&L provided an external subject matter review.

ICALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO.4 1PAGE NO.3 of 117 REVISION HISTORY Revision Revision Description 0 Original issue. I Removed the credit of FRVS recirculation charcoal filter efficiencies, 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. 2 Revised to assess radiological impact of increased core thermal power level of 3,917 MWt. As of 12/07/2005, the EPU project decided to adopt the AST analysis performed for the increased core thermal power level for the current design and licensing bases because it conservatively bounds the EPU project design. Section 8.2 indicates that the proposed increases in the EAB, LPZ, and CR doses and the total doses are less than the corresponding minimal dose increases and applicable regulatory allowable limits as defined in the 10 CFR 50.59 rule. The implementation or cancellation of the proposed core thermal power related DCP would not have any adverse impact on this analysis. Some of the design inputs are taken from documents that support higher core thermal power operation. If the HCGS license is not amended for the proposed increased power level, these design inputs would become conservative assumptions without having any adverse impact on the validity of this analysis. 3 Analysis is revised to evaluate the radiological impact of increases in the primary containment isolation valves (PCIVs) closure time to 120 seconds and ESF leak rate to 2.85 gpm. The MSIV leakage aerosol deposition model is revised in a conservative manner consistent with the regulatory guidance to create the dose margin for these changes. The revised aerosol deposition model is consistent with the NRC approved models documented in recent BWR AST license amendments. This is a General Revision (no revision change bars) to reformat for compliance with EXELON calculation procedure, resulting in some Revision 2 text rolling from one page to the next page. 4 Analysis is revised to delete the evaluation of the increased PCIV closure time and to include other modeling changes listed in Table 1E.

CALCULATION NO. H-1-ZZ-MDC-1880 I REVISION NO. 4 [ PAGE NO. 4 of 117 TABLE OF CONTENTS Design Analysis Cover Sheet I Table of Contents 4 1.0 Purpose 5 2.0 Methodology 5 3.0 Acceptance Criteria 19 4.0 Assumptions 20 5.0 Design Inputs 24 6.0 Computer Codes and Compliance With Regulatory Requirements 32 7.0 Calculations 34 8.0 Results Summary 51 9.0 Conclusions 58 10.0 References 59 11.0 Tables 64 12.0 Figures 97 13.0 Affected Documents 104 14.0 Attachments 104

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO. 4 PAGE NO. 5 of 117

1.0 PURPOSE

The purpose of this calculation is to evaluate the post-LOCA Exclusion Area Boundary (EAB), Low Population Zone (LPZ), and Control Room (CR) doses for Hope Creek Generating Station (HCGS) using conservative as-built design inputs and assumptions with an assumed core thermal power level of 3,917 MWt. 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. The following design basis post-LOCA release paths are analyzed:

1. Containment Leakage.

2. Engineered Safety Feature (ESF) Leakage.

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

This calculation is revised in response to Notification 20470663 which requires that the evaluation of the increased Primary Containment Isolation Valve (PCIV) closure time of 120 seconds be removed from the analysis. The calculation is also being revised to

• eliminate the 50% reduction in containment leakage and MSIV leakage after 24 hours,
• to change the number of main steam pipe segment volumes in the MSIV leakage analysis RADTRAD model from two volumes in each of the intact and failed main steam line pathways to one volume in each of the intact and failed main steam line pathways,
• to change the percentile aerosol deposition settling velocities from 5 0th %/ 30 th % that was used in the two volume RADTRAD model to 4 0 th %,
• to change the approach for calculating elemental iodine removal in the main steam lines, and
• to reduce the control room envelope unfiltered inleakage rate to 300 cfm.

The resulting revision incorporates elements of the MSIV leakage model that were used in Revision 2. A sensitivity study was performed to address the addition of 12 Co-60 ITAs to the HCGS core - See Attachment 14.2. 2.0 METHODOLOGY The design basis LOCA 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 to demonstrate compliance with 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. Changes in input parameters and methodologies between Revisions 2, 3, and 4 of this calculation are identified in Table lE. 2.1 Radiation Source Terms: The reactor core radionuclide inventory is based on a maximum power level of 3917 MWt, which includes 2% instrument uncertainty (Assumption 4.2). Radiation source term radionuclides used in this

CALCULATION NO. H-1-ZZMDC-1880 REVISION NO. 4 PAGE NO.6 of 117 evaluation are those used in Alternative Source Term analyses and are identified in Tables IA through ID. Radionuclide inventories (Ci/MWt) presented in Tables I A and lB are bounding inventories for a discharged bundle and the radionuclide inventories provided in Tables IC and ID are average inventories over the reactor core. Since this is a LOCA evaluation which involves the release of radionuclides from the entire core, the average radionuclide inventories from Tables I C and ID are used in this evaluation. 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 composite average core inventory is shown in Table ID. The RADTRAD Nuclide Inventory File (NIF) HEPULOCAI1def.txt is developed based on the uprated core inventory in Table ID 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. HCGS intends to include 12 Isotope Test Assemblies (ITAs) in the reactor core. The ITAs will contain Co-59 targets which will be irradiated in the reactor core in order to produce Co-60. GE-Hitachi Report NEDC-33529P and RAI#17 of PSEG Letter LR-N10-0163 provide an analysis of the impact of the introduction of the 12 ITAs on LOCA dose evaluations. The impact on doses was determined to be negligible for the control room, EAB, and LPZ doses. The same analysis was re-performed using the design inputs and assumptions in Sections 4.0 & 5.0 of this Revision 4 of the calculation (see Attachment 14.2). The same result of negligible impact was obtained. 2.1.1 Activity Transport In Primary Containment (Table IE. Item I-1): The average core inventory (Ci/MWt) listed in Tables IC and ID is released into the containment at the release timing and fractions shown in RG 1.183 Tables 3 and 4 (Ref. 10.1, RGPs 3.2 and 3.3). The RADTRAD default Release Fraction and Timing (RFT) file BWRDBA starts the gap release phase at time t= 0 sec after onset of a LOCA without 2 minutes delay (Ref. 10.1, Table 4). The same BWR_DBA RFT file was used in the previous revisions of this analysis. 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 RG 1.183 RGP 3.5 (Assumption 4.5). 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). In accordance with Assumption 3.1 of Appendix A to R.G. 1.183, the radioactivity released from the fuel is assumed to mix instantaneously and homogeneously throughout the free air volume of the primary containment (drywell) as discussed below. 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 5.3.2.10. Radioactivity is released from the core and diluted into the drywell air volume. Following the initial blowdown of the reactor pressure vessel, the steaming in the RPV carries fission products to the containment. When core cooling is restored, the fuel damage is terminated. The steam and the ESF flows carry any remaining fission products from the vessel, through the break, to the primary containment and provide new steam flow for rapid drywell- suppression air space mixing. After 2 hours the containment drywell and suppression chamber air volumes are expected to become well mixed due to a very high flow established between drywell and wetwell as a result of steaming and condensing phenomenon (Ref. 10.22, Table 2) and thus after 2 hours, airborne activity dilution is credited in the combined air volumes of the drywell and suppression chamber (Table IE, Item 1-1). Taking credit for dilution of containment airborne activity within the drywell and suppression chamber combined air volumes over the remaining course of the accident (after 2 hours) is in accordance with guidance provided in Section 3.1 of Appendix A to R.G. 1.183. In Revision 2 of this calculation, radionuclide activity released into containment was assumed to mix instantaneously within the drywell and

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO. 4 1 PAGE NO. 7 of 117 suppression chamber air volumes. Confining the airborne activity to the drywell air volume during the first 2 hours is conservative because it increases the amount of airborne activity released from containment during the first 2 hours and increased activity releases result in larger doses. 2.1.2 Reduction In Airborne Activity Inside Containment Aerosol Deposition The gravitational deposition of aerosols from the containment atmosphere is credited by using the RADTRAD "POWERS MODEL" with 10 percentile uncertainty distribution resulting in the lowest removal rate of the aerosols from the containment. Crediting the reduction in aerosol airborne activity from the containment atmosphere by using the RADTRAD "Powers Model" with 10 percentile uncertainty is in accordance with the methodology described in Section 3.2 of Appendix A to R.G. 1.183 and in AEB-98-03. Suppression Pool Scrubbing Iodine removal by suppression pool scrubbing is not credited because the timed release of activity associated with the Alternative Source Term methodology results in the bulk of core activity being released to containment well after the initial mass and energy release. Natural Deposition on Containment Surfaces (Table IE, Item 1-2) In accordance with Section 3.2 of Appendix A to R. G. 1.183, reduction in containment airborne radioactivity by natural deposition within containment may be credited using the models described in SRP 6.5.2, even if no credit is taken for containment sprays (Table 1E, Item 1-2). The Decontamination Factor (DF) of elemental iodine is based on the Standard Review Plan (SRP) 6.5.2 guidance and is limited to a DF of 200 (see Section 7.11) (Ref. 10.41, page 6.5.2-12). The SRP 6.5.2 calculation of the elemental iodine removal rate is based on a minimized wetted surface area that conservatively results in a smaller elemental removal coefficient and a longer time to reach an elemental iodine decontamination factor (DF) of 200 (see Section 7.10). This longer time to reach the DF allows the elemental iodine to remain airborne in the drywell atmosphere for a longer time prior to release to the environment via containment and MSIV leakage. This results in a larger release and a higher dose. The drywell wetted surface area is conservatively minimized by crediting 25% of the drywell lining surface, and 50% of the major equipment and structure surfaces, and then by applying a 25% reduction to the estimated 2surface area. The resultant modeled surface area of 33,200 ft is less than half of the available 69,126 ft 2 drywell wetted surface area. The removal of elemental iodine by the wetted surface area is consistent with the guideline provided in Standard Review Plan (SRP) 6.5.2 (Ref. 10.41) and RG 1.183 (Ref. 10.1). The suppression pool water pH is greater than 7 (Design Input 5.4.4). A pH greater than 7 inhibits the iodine that deposits on containment surfaces from re-evolving back into containment atmosphere during containment spray recirculation. The RADTRAD code calculates the elemental and organic iodine atoms in the drywell atmosphere. The following procedure is established to calculate the cutoff time of the elemental iodine removal by wall deposition inside the drywell:

1. The isotopic elemental iodine atoms are calculated using the atoms/curie relationship established in Table 8.

2. The initial isotopic elemental iodine activity in the drywell is determined based on the RG 1.183 (Section 3.2, Table 1, and Section 3.5), which is 4.85% of the total 30% iodine released in the drywell (Ref. 10.1, Table 1) (see Table 9).

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO. 4 PAGE NO. 8 of 117

3. The initial isotopic elemental iodine atoms in the drywell are totaled and then divided by the Decontamination Factor of 200 to determine the elemental iodine atoms expected to be in the drywell when the DF of 200 is reached, which is 3.752E+20 atoms (Table 9).

4. The containment leakage case is analyzed in RADTRAD Run CUTOFF.oO using an elemental iodine removal rate of 3.16 hr-1 for the first 2.0 hours of the accident, and then an elemental iodine removal rate of 1.74 hf-1 for the remainder of the accident as calculated in Section 7.11. This RADTRAD run provides the elemental iodine atoms in the containment at different time intervals.

RADTRAD output CUTOFF.o0 indicates the drywell elemental iodine atoms reach a value of 3.8056E+20 at 4.02 hrs and 3.7400E+20 at 4.03 hrs. This means that an elemental iodine DF of 200 is reached in the drywell at 4.03 hrs. Therefore, elemental iodine wall removal is not credited in the analysis beyond 4.0 hrs. 2.2 POST-LOCA CONTAINMENT LEAKAGE PATHWAY (TABLE 1E, ITEMS 111-1 AND VI-1) The post LOCA containment leakage model is shown in Figure 1. The time dependent radionuclide source term airborne in containment and available for leakage is that described in Section 2.1 of this evaluation with credit taken for dilution (Section 2.1.1) and reduction of airborne activity (Section 2.1.2). The containment is assumed to leak at the technical specification leak rate of 0.5 volume percent per day over the course of the accident, i.e., 720 hours (Design Input 5.3.2.5). No credit is taken for a reduction in containment leak rate (e.g., at 24 hours post LOCA or subsequently during the LOCA duration) due to a reduction in containment pressure with time (Table IE, Item Ill-1). Keeping the containment leakage rate at the technical specification limit for the duration of the accident is conservative since it maximizes the amount of radioactivity released which maximizes doses. Primary containment leakage is into the reactor building, however, for the first 375 seconds primary containment leakage is released directly to the environment with no credit for holdup or filtration. The time of 375 seconds is the draw down time (Design Input 5.3.2.6), i.e., the time it takes to bring the reactor pressure down to 0.25-inch water gauge negative pressure relative to adjacent areas as defined in technical specifications (Ref. 10.1 RGP A.4.2). After 375 seconds, primary containment leakage is via the reactor building through the FRVS filter system. The FRVS recirculation system circulates reactor building air at a recirculation flow rate of 108000 cfm (Design Input 5.3.2.16). The volume of the reactor building is 4.OOE+06 cubic feet (Design Input 5.3.2.9) and thus the FRVS recirculation system recirculates reactor building air at a rate of 1.62 reactor building volumes/hr (i.e., 108000 cfm x 60 min/hr x (4.OOE+06 cu ft)-'). Although the FRVS provides good mixing of airborne activity within the reactor building, for conservatism, credit is taken for mixing and dilution of activity in only 50% of the reactor building volume (Ref. 10.1, RGP A.4.4). To account for 50% mixing in the reactor building, the FRVS vent system exhaust rate to the environment is doubled (Design Input 5.3.2.15). The FRVS ventilation system exhaust rate varies with time as shown by the equation in Design Input 5.3.2.12. Design Input 5.3.2.15 provides the FRVS exhaust flow rates at 100% and 50% mixing. Airborne activity in the reactor building is reduced by both the FRVS recirculation and FRVS vent filtration systems before it is released to the environment. FRVS exhaust filter and FRVS recirculation filter efficiencies are presented in Design Inputs 5.3.2.13 and 5.3.2.14, respectively. Containment Purge During Normal Operation The Bases for HCGS Technical Specification Section 3.6.1.8 (Ref. 10.6.18) states that the use of the drywell and suppression chamber purge exhaust lines for pressure control during plant operational conditions 1, 2, and 3 is unrestricted provided 1) only the inboard purge exhaust isolation valves on these lines and the vent valves on the 2-inch vent paths are used and 2) the outboard purge exhaust isolation valves remain closed. It is also required that the design of the purge supply and exhaust isolation valves and the 6" nitrogen supply valves meet the requirements of Branch Technical Position

CALCULATION NO. H-1-ZZ-MDC-1880 I REVISION NO. 4 1 PAGE NO. 9 of 117 CSB 6-4, "Containment Purging During Normal Plant Operations". Item B. 1 .F of Branch Technical Position 6-4 of NUREG-0800 (Ref. 10.56) states that the purge system isolation valve closure times, including instrumentation delays, should not exceed 5 seconds to facilitate compliance with 10 CFR 100 for offsite radiological consequences. The drywell and suppression chamber purge supply and exhaust valves and the 6" nitrogen supply valves are shown on HCGS Containment Atmosphere Control P&ID M-57-1 (Ref. 10.53). The relevant valves are drywell purge exhaust isolation valves HV-4950 and HV-4952, drywell purge supply isolation valves HV-4956 and HV-4979, suppression chamber purge supply isolation valves HV-4980 and HV-4958, suppression chamber purge exhaust isolation valves HV-4962 and HV-4964, and nitrogen purge isolation valve HV-4978. Table 3.6.3-1 of the HCGS Technical Requirements Manual (Ref. 10.54) indicates that the maximum isolation time for each of these valves is 5 seconds. HCGS Technical Specification Section 3.6.1.8 (Ref. 10.6.18) states that the drywell and suppression chamber purge system, including the 6 inch nitrogen supply line, may be in operation for up to 500 hours each 365 days with the supply and exhaust isolation valves in one supply line and one exhaust line open. The Bases to Technical Specification 3.6.1.8 states that the 500 hour/365 day limit for operation of the purge valves and the 6" nitrogen supply valve during plant operational conditions 1, 2, and 3 is intended to reduce the probability of a LOCA occurrence when the applicable combination of the above valves are open. A dose analysis was performed that concludes that the dose associated with a containment purge while operating, concurrent with LOCA is negligible (UFSAR 1.14.1.7) compared to the dose due to the release of design basis LOCA sources. The negligible purge dose is based on limiting the purge isolation valve closure time to 5 seconds and so the radiation source term released via the purge system is reactor coolant. In light of the previous discussion, calculation of radiological consequences is not required due to a normal containment purge coincident with a LOCA (Table 1E, Item VI- 1). Containment Purge During LOCA Section 7 of Appendix A to R.G. 1.183 states that if post LOCA primary containment purging is performed as a combustible gas or pressure control measure or if primary containment purging is required within 30 days following a LOCA, then radiological consequences should be analyzed. It further states that if the containment purging capabilities are maintained for purposes of severe accident management and are not credited in any design basis analysis, then radiological consequences need not be evaluated. The HCGS containment is not purged for combustible gas or pressure control measure within 30 days following a LOCA (Assumption 4.6.3). Therefore, in accordance with guidance provided in Section 7 of Appendix A to R.G. 1.183, radiological consequences associated with a purge release during LOCA conditions need not be evaluated (Table I E, Item VI-1). 2.3 Post-LOCA ESF Leakage Pathway (Table 1E, Item IV-i): 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 the 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 Reactor Building. The ESF leakage rate is 2.85 gpm (Design Input 5.4.2). The ESF leakage rate was increased from 1 gpm (in Revision 2 of this calculation) to 2.85 gpm in Revision 3 of this calculation to facilitate operational flexibility by providing additional operating margin for ESF leakage (Table 1E, Item IV- 1). This ESF leak rate is doubled (Ref 10.1, Section A.5.2) and assumed to start at time t=0.0 minute after onset of a LOCA. Ten percent of the iodine in the suppression pool water that leaks becomes airborne

CALCULATION NO. H-l-ZZ-MDC-1880 REVISION NO. 4 PAGE NO. 10 of 117 (Design Input 5.4.6). 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. ESF leakage is into the reactor building. As indicated in Section 2.2 of this calculation, for containment leakage, airborne radioactivity released due to ESF leakage is released directly to the environment for the first 375 seconds, i.e., during draw down. After 375 seconds, credit is taken for mixing in the reactor building volume and for airborne activity removal by the FRVS recirculation and exhaust filter systems. The EAB, LPZ, and CR TEDE doses that result from ESF leakage are shown in the Section 8.0 of this calculation. 2.3.1 Suppression Pool Water 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 and 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 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). 2.4 Post-LOCA MSIV Leakage Pathway (Table 1E, Items V-1, V-2): The post-LOCA MSIV Leakage model is shown in Figures 3 and 4. The four main steam lines, which penetrate the primary containment, are automatically isolated by the MSIVs in the event of a LOCA. There are two MSIVs on each steam line, one inside containment and one outside containment. The MSIVs are functionally part of the primary containment boundary and design leakage through these valves provides a leakage path for fission products that bypass the secondary containment and enter the environment as a ground-level release. MSIV leakage is postulated to be release to the environment through the MSIV failed steam line and one of the three remaining intact steam lines. The main steam isolation valves (MSIVs) are postulated to leak at a total design leak rate of 250 scfh at 50.6 psig. This is the maximum allowable technical specification value (250 scfh combined through all 4 main steam lines per Design Input 5.5.1). Measured leakage through any one steam line cannot exceed 150 scfh. In order to maximize dose results, the leak rate of 150 scfh is applied to the main steam line with the failed MSIV (Design Input 5.5.2) and the leak rate of 100 scfh is applied to the intact main steam line (Table IE, Item V-2). Use of the maximum technical specification leak rate is in accordance with Assumption 6.2 of Appendix A to R.G. 1.183. The time dependent radionuclide source term airborne in containment and available for leakage via the main steam lines is that described in Section 2.1 of this evaluation with credit taken for dilution (Section 2.1.1) and reduction of airborne activity (Section 2.1.2). As indicated in Section 2.2, no credit is taken for a reduction in containment leak rate due to a reduction in containment pressure. This is also true for the MSIV leak rate, i.e., no credit is taken for a reduction in the MSIV leak rate due to a reduction in containment pressure at 24 hours post LOCA or subsequently during the LOCA duration (Table IE, Item V-1). MSIV leakage is modeled in this revision using a single main steam pipe volume on the failed main steam line and a single main steam pipe volume on the intact main steam line. Revision 2 of this evaluation also used a single main steam pipe volume for the failed main steam line and for each intact main steam line, whereas Revision 3 used two main steam pipe volumes on both the failed main steam line and on one intact main steam line (Table lE, Item V-5). Use of a single main steam line pipe

CALCULATION NO. H-1-ZZ-MDC-1880 I REVISION NO. 4 1 PAGE NO. 11 of 117 volume in lieu of two pipe volumes is conservative in that dose consequences of the release are larger for a single main steam pipe volume than they are when multiple pipe volumes are used. Dose consequences are smaller when multiple main steam pipe volumes are considered in series because there is a compounding of the radioactivity removal efficiencies for the mechanisms that remove airborne radioactivity from the pipe volume, such as aerosol deposition. 2.4.1 MSIV Leak Rates (Table IE, Items V-2 and V-8) A total of 250 scth MSIV leakage is assumed to occur as follows: (1) 150 scfh through the shortest steam line. This line is modeled as having the failed inboard MSIV. The deposition of aerosol and removal of elemental iodine activities are not credited in the steam line between the RPV nozzle and the inboard MSIV because this section of pipe is assumed to be broken and exposed directly to the drywell atmosphere. The deposition of aerosols and removal of elemental iodine are conservatively credited only in the horizontal pipe between the inboard MSIV and turbine stop valve (TSV) for 0-96 hrs only. A removal time between 0 and 96 hours was also used in Revision 3 of this calculation whereas a removal time of 0 to 30 days was used in Revision 2 of this calculation. The shorter removal time means less of the aerosols and elemental iodine will be removed which results in greater dose consequences which is conservative. (2) 100 scfh through shortest of the three intact steam lines. The deposition of aerosol and removal of elemental iodine activities are conservatively credited only in the horizontal pipe segments between the RPV nozzle and TSV for 0-96 hrs only. Revision 2 modeled 50 scfh in each of two intact steam lines. (3) 0 scfh through second shortest intact steam line. (4) 0 scfh through third shortest intact steam line. Since the shortest steam line is allotted the maximum allowed leakage, it is assured that leakage through any other line is bounded. Use of the shortest main steam line as the failed steam line and use of the shortest of the three intact steam lines as the intact steam line minimizes pipe surface areas for aerosol deposition and elemental iodine removal which minimizes the amount of aerosols and elemental iodine removed which is conservative. No credit is taken for a reduction in the MSIV leakage due to a reduction in containment pressure with time. Keeping the MSIV leakage rate at the technical specification limit for the duration of the accident is conservative since it maximizes the amount of radioactivity released which maximizes doses. 2.4.2 Main Steam Line Flow Model (Table IE, Item V-4) A well mixed flow model is used to model the transport of radioactivity within the main steam pipe volumes. Revision 3 to this calculation also used a well mixed flow model, however, a plug flow model that took credit for holdup of radioactivity within the main steam pipe volume, was used in revision 2 (Table 1E, Item V-4). Credit for holdup in the main steam pipe volume is not applicable when the well mixed flow model is used. The basis for using a well mixed flow model instead of the plug flow model is provided in R.G. 1.183 (Ref. 10.1) and AEB-98-03 (Ref. 10.22). Assumption 6.3 of Appendix A to R.G. 1.183 discusses

CALCULATION NO. H-1-ZZ-MDC-1880 I REVISION NO. 4 PAGE NO.12 of 117 reduction in radioactivity in steam system piping by deposition and plateout. It states that the model used to address reduction in the amount of released radioactivity by deposition and plateout should generally be based on the assumption of well mixed volumes, but other models such as slug flow may be used if justified. A discussion in AEB-98-03 states that for the same leak rate into the main steam line, plug flow is expected to result in less offsite release than the well-mixed flow, because the concentration of the material released to the environment is at the concentration of the material in the plug at the end of the pipe. The text in AEB-98-03 goes on to state that the NRC staff believes that, at this time, a well mixed model is more appropriate than the plug flow model for settling along the main steam lines. Since the well mixed flow model results in greater radioactivity releases from the main steam pipe volumes for the same leak rate, use of the well mixed flow model will result in larger doses and is therefore conservative. 2.4.3 Determination of MSIV Volumetric Leak Rates (Table 1, Item V-3) The total MSIV leakage from all main steam lines is assumed to be 250 scfh measured at 50.6 psig, with a maximum of 150 scfh from any one of the 4 main steam lines. The total MSIV leak rate of 250 scfh is converted using the ideal gas law to determine the actual leakage rate (cfh) based on the post-LOCA peak temperature and pressure in Section 7.2 (Table IE, Item V-3). The actual MSIV leak rate from the drywell atmosphere into the main steam pipe volumes is smaller than the measured leak rates due to the combined effects of accident condition compression (due to the high pressure) and expansion (due to the high temperature). Use of the actual MSIV leak rate based on drywell LOCA conditions is supported by RIS-2001-19 (Ref. 10.57). Section 3.f of Reference 10.57 states that with regard to volumetric flow rates specified as Limiting Conditions of Operation (LCOs), the density used should be consistent with the density that is assumed in the surveillance procedure that demonstrates compliance with the LCO. Since the peak drywell pressure is incorporated into the MSIV leakage surveillance requirement (Technical Specification LCO 3.6.1.2.c, Ref. 10.6.17), adjusting for actual drywell conditions is appropriate. MSIV leak rates from the main steam line pipe volumes to the environment through the TSVs are conservatively calculated in Section 7.2 and listed in Table 7, using the Ideal Gas Law and drywell post-LOCA peak pressure and temperature conditions in the pipe volume and standard atmospheric conditions for the environment. To account for the assumed mixing between the wetwell and drywell after 2 hours and the resulting activity dilution, the flow rate through the MSIVs is reduced by the ratio of the drywell volume to the total volume at two hours (Section 7.2). MSIV leakage flow rates from the drywell that are calculated in Revision 3 of this evaluation are also calculated using post LOCA drywell temperature and pressure. Revision 2 of this calculation directly used the measured MSIV leakage values that represent standard atmospheric pressure and temperature conditions. Adjusting the measured MSIV leak rates for LOCA drywell pressure and temperature conditions to quantify MSIV leak rates for Revisions 3 and 4 of this evaluation represents an adjustment of input parameters to reflect actual plant conditions. A comparison of the MSIV leak rate reductions between Revision 2 to Revision 4 for both the MSIV failed and intact steam lines is shown in the following table:

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO. 4 1 PAGE NO. 13 of 117 MSIV Leakage from Drywell Post-LOCA Revision 2 Revision 4 Time Interval MSIV Failed 2 Intact MSIV Failed 1 Intact Line Lines Line Line A B C D (hr) (scfh & cfm) (scfh & cfm) (scfh/cfh & cfm) (scflicfh & cfm) 0-2 150 50 150/48.47 100/32.31 2.5 0.833 0.808 0.539 2-24 150 50 150/26.77 100/17.84 2.5 0.833 0.446 0.297 24-720 75 25 150/26.77 100/17.84 1.25 0.417 0.446 0.297 A From H-1-ZZ-MDC-1880, Rev 2, Section 5.5.2 A From H-1-ZZ-MIDC-1880, Rev 2, Sections 5.5.3 and 5.5.4 C and D from Table 7 A review of the MSIV leak rate information in the above table indicates that there is a substantial reduction in the post-LOCA MSIV cfm leakage rates in both the MSIV failed and intact lines, which reduces the dose contribution resulting from the MSIV leakage release path without accounting for the aerosol deposition. 2.4.4 Aerosol Deposition in Main Steam Piping (Table IE, Item V-8) 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 AEB-98-03 (Ref. 10.22), which uses the Monte Carlo distribution of aerosol settling velocity in well mixed flow. Assumption 6.5 of Appendix A to R.G. 1.183 allows for a reduction in MSIV releases that is due to holdup and deposition in main steam piping downstream of the MSIVs if the piping is capable of performing its safety function during and following a safe shutdown earthquake. The analysis in Section 7.4 of this evaluation takes credit for the aerosol deposition and elemental iodine removal in piping upstream and downstream of the outboard MSIVs because the steam lines from the RPV nozzles to the TSVs are seismically designed and supported for the safe shutdown earthquake (Ref. 10.26 and 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 aerosol deposition removal efficiencies for the main steam lines are determined based on the methodology in Appendix A of AEB-98-03 (Ref. 10.22) as shown in Tables 2 through 6 using the information in Sections 7.3 and 7.4. Credit is taken for aerosol removal from main steam piping via deposition for times between 0 and 96 hours post LOCA (Table IE, Item V-8). Main Steam Pipe Segment Lengths for Aerosol Deposition (Table 1E, Items V-5 and V-6) The failed main steam line consists of a single well mixed volume between the inboard MSIV and the TSV. The intact main steam line consists of a single well mixed volume between the RPV and the TSV. In Revision 3 of this calculation, the failed main steam line consisted of 2 well mixed pipe volumes. One volume was between the inboard and outboard MSIV and the second volume was between the outboard MSIV and the TSV. The intact main steam line in Revision 3 of this calculation also consisted of 2 main steam pipe volumes. One volume was between the RPV nozzle and the outboard MSIV and

CALCULATION NO. H-1-ZZ-MDC-1880 I REVISION NO. 4 PAGE NO.14 of 117 the second volume was between the outboard MSIV and the TSV. The failed main steam line pipe volume in Revision 2 of this calculation consisted of one plug flow volume that extended from the RPV nozzle to the TSV, while conservatively neglecting the main steam line length between the inboard and outboard MSIVs. The intact main steam pipe volumes in Revision 2 of this calculation for the 2 intact main steam lines extended from the RPV nozzle to the TSV. The main steam piping layouts in the drywell through main steam tunnel to the TSV are shown in the piping isometric drawings in References 10.12 and 10.21 with the piping parameters. The review of these isometric drawings indicates that the steam header A and the steam header D share the shortest horizontal pipe length between the RPV nozzle and inboard MSIVs, which provides the minimum horizontal pipe surface area and consequently results in the least aerosol deposition. The main steam piping volumes and horizontal pipe surface are calculated in Section 7.3 and listed in Tables 2 through 4. Main Steam Pipe Surface Areas for Aerosol Deposition (Table IE, Item V-6) Main steam pipe aerosol deposition surface areas for the failed and intact main steam lines are calculated in Sections 7.3.2 and 7.3.3 of this calculation, respectively. For conservatism, deposition of aerosols is only credited in horizontal pipe segments. Gravitational force naturally removes the airborne aerosol particles in the main steam piping during their migration through the pipe to the atmosphere. The horizontal main steam piping projected surface area (Diameter x Length) provides a favorable condition for aerosol deposition (Table IE, Item V-6). Only considering aerosol deposition in horizontal pipe segments and calculating a settling area as D x L, is in accordance with Section 2 of Reference 10.55. Aerosol deposition surface areas in Revision 3 of this calculation were also calculated for a projected surface area by applying the formula of D x L to horizontal pipe surface areas. In the Revision 2 analysis, the total pipe surface area (n x D x L) was used for aerosol deposition. Use of the total pipe surface area instead of the projected surface area overestimates the deposition surface since aerosol deposition is a result of gravitational force. Use of the projected surface area in lieu of the total pipe surface area is conservative since it results in a smaller surface area for aerosol deposition and thus removal of a smaller fraction of the aerosol activity. Percentile Settling Velocity for Aerosol Deposition (Table I E. Item V-7) AEB-98-03 (Ref. 10.22) states that given the conservatisms associated with using a well mixed model for the entire length of the pipe and a number of additional conservatisms inherent in the piping deposition analysis, use of a 10 th percentile settling velocity with a well mixed model is not appropriate. It goes on to state that given the conservatism of the well mixed assumption, it is acceptable to use median values for percentile settling velocity as opposed to more conservative values. The rate constants (ks) for different piping segments in the MSIV leakage release paths are calculated in Table 5 using a 4 0 1h percentile aerosol settling velocity in piping segments upstream and downstream of the outboard MSIV in both the failed and intact main steam lines (Ref. 10.22, Appendix A, Table A-i) and applicable horizontal settling areas and volumes from Table 4 (Table IE, Item V-7). The aerosol removal efficiencies due to gravitational depositions on the horizontal pipe surfaces are calculated using the mass balance equation for the well mixed volumes in Section 7.4 and listed in Table 6. The 4 0 th percentile settling velocity is smaller than the mean (i.e., 5 0 th) percentile settling velocity recommended in AEB-98-03. Use of the smaller settling velocity means there is more aerosol activity available for release which results in larger dose consequences and is therefore conservative. A 401h percentile settling velocity was also used in Revision 2 of this calculation to determine aerosol deposition removal rate constants for the failed and intact main steam lines. The main steam pipe model

CALCULATION NO. H-1-ZZ-MDC-1880 I REVISION NO. 4 PAGE NO. 15 of 117 used in Revision 3 of the calculation consisted of 2 pipe segment volumes in series in both the failed and intact main steam lines. Percentile settling velocities of 5 0 th percentile and 3 0 th percentile were used for the 2 pipe volumes in both the failed and intact main steam lines. The 3 0 th percentile settling velocity was applied to the 2 nd, downstream pipe volume and is smaller to account for the reduction in the size of the aerosol particles in the downstream volume. The overall effect of changing from a two compartment model with a 5 0 th percentile settling velocity in the first compartment and a 3 0 th percentile settling velocity velocity in is athe second larger compartment release to a single which results compartment in larger doses. model with a 401h percentile settling 2.4.5 Elemental Iodine Removal (Table IE, Items V-8 and V-9) The reduction in elemental iodine activity in the MSIV leakage is calculated in Section 7.4.2 using the staff recommended guidance for an acceptable method as documented in Reference 10.52 (Reference A-9 of RG 1.183). The natural removal efficiencies for elemental iodine in each steam line volume are calculated in Tables IF through IM using the J.E. Cline model (Table IE, Item V-9). The elemental iodine removal efficiencies calculated in Tables IF through IM are as previously approved by the NRC in the Hope Creek AST license amendment (Ref. 10.49). The information in these tables are revised due to minor changes in the surface areas and volumes. The J. E. Cline method for calculation of elemental iodine removal from the main steam pipe volumes was also used in Revision 2 of this calculation. Revision 3 of this calculation applied a factor of 2 reduction for elemental iodine in each main steam line volume. The factor of 2 reduction is in accordance with the methodology prescribed in Appendix B of AEB-98-03 (Ref. 10.22). Credit is taken for removal of elemental iodine activity from the main steam piping for times between 0 and 96 hours post LOCA (Table 1E, Item V-8). 2.4.6 Parametric Study to Validate Aerosol Deposition Model Parametric studies are not needed to validate the aerosol deposition model used in the MSIV leakage analysis. (1) The elemental iodine removal efficiencies used in Revision 4 (Tables IF through IM) have been previously reviewed by the NRC staff and approved in the Hope Creek AST license amendment (Ref. 10.49). The post-LOCA temperature profile used for these elemental iodine removal efficiencies remains bounding for EPU (Section 5.3.1.9). (2) Revision 4 uses one well mixed volume node for the MSIV failed line and for the intact line. This model was previously reviewed by the NRC staff and approved in the Hope Creek AST license amendment (Ref. 10.49). (3) In Revision 4, the 4 0 th percentile aerosol settling velocity is used for the one well mixed volume in the MSIV failed line and in the intact line, which is conservative with respect to AEB 98-03, which states that the use of the 50th percentile aerosol settling velocity is appropriate. The use of the 40th percentile aerosol settling was previously reviewed by the NRC staff and approved in the Hope Creek AST license amendment (Ref. 10.49).

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO. 4 PAGE NO. 16 of 117 2.5 Control Room Model The post-LOCA control room RADTRAD nodalization is shown in Figure 5 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
• Post-LOCA airborne cloud external to CR
• Post-LOCA containment shine to CR 9 Post-LOCA CREF filter shine 2.5.1 Post-LOCA Airborne Activity Inside CR The post-LOCA radioactive releases from various sources are discussed in Sections 2.2 through 2.4 above and shown in Figure 5. The activity releases from the various sources are diluted by the atmospheric dispersions and carried to the CR air intake. The atmospheric dispersion factors are shown in Sections 5.6.9 and 5.6.11 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.

2.5.2 Control Room Unfiltered Inleakage (Table IE, Item 11-1) The maximum CR inleakage is measured to be 155 +/- 10 cfm in the recently performed (year 2009) Tracer Gas Test (Ref. 10.46, Table 19). In the last two Tracer Gas Tests, all CR inleakage was consistently measured to be filtered. Although all CR inleakage is filtered, the analyses in Revision 2 and Revision 3 assumed 350 cfm unfiltered inleakage, which is conservative. This Revision 4 analysis assumes the CR unfiltered inleakage is 300 cfm, including 10 cfm for ingress/egress. Use of 300 cfm in Revision 4 reduces the control room dose. This assumption remains conservative with respect to the fact that all CR inleakage is filtered. 2.5.3 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 at EL 155'-3" and 1'-10-1/2" roof at 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.

SCALCULATION NO. H-1-ZZ-MDC-1880 I REVISION NO. 4 1 PAGE NO.17 of 117 2.5.4 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. 2.5.5 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, and 10.39). The CR operating floor is located at elevation 137'-0" (Ref. 10.28c). The concrete floor at EL 155'-3" is 1 foot 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 the charcoal/HEPA filter. The review of the CREF unit locations (Ref. 10.39) and CR area normally occupied by the CR operator (Ref. 10.30) indicates that the CR area below the CREF units are not occupied during routine operation. The CREF unit 1AVH400 is relatively closer to the normally occupied CR area compared to CREF unit 1BVH400 (Refs. 10.29, 10.30, and 10.39). Using full size drawings (Refs. 10.30 and 10.39), Section 7.6.4 determines that the slant distance and slant angle between the center line of CREF unit 1AVH400 and the horseshoe control panel and console are 30.99 feet and 27.38', respectively. Based on this slant angle, Section 7.6.4 determines that the slant concrete shielding provided by the 1'-6" concrete ceiling and used in the MicroShield model is 3.26 feet. 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' (H) x 4' (W) by summing all of the charcoal filter trays within a filter housing as shown in Figure 6, which also shows the dose point location. The post-LOCA aerosol buildup on the HEPA filter and the iodine buildup on the charcoal filter are calculated as follows. 2.5.5.1 Post-LOCA Iodine and Aerosol Activity On CREF Charcoal/HEPA Filter - Containment Leakage The RADTRAD3.02 code calculates the cumulative elemental and organic iodine atoms and the aerosol mass deposited on the CR recirculation charcoal/HEPA filters. The CREF intake filter iodine and aerosol activities are calculated in Section 7.6.1 for the containment leakage. The relationship between the aerosol mass and activity is established in Table 11 based on the information obtained from RADTRAD run HEPU300CLOO.oO. The aerosol mass deposited on the CREF HEPA recirc filter is calculated by the RADTRAD code for the duration of the accident. Knowing the CR intake and recirc filtration flow rates, the relationship can be established to calculate the aerosol mass deposited on the intake HEPA filter as shown in Section 7.6.1. The total aerosol mass deposited on the CREF HEPA filter due to the containment leakage is calculated, which is used with the aerosol mass/activity relation established in Table 11 to calculate the aerosol isotopic activities deposited on the CREF HEPA filter. This aerosol mass is added with the aerosol mass from the MSIV leakage in Section 7.6.3A and then converted into the isotopic aerosol activities in Table 12 using the atom/activity relation in Table 11. The total (elemental + organic) iodine atoms deposited on the CREF charcoal filter due to the containment leakage are calculated in Section 7.6.1. The iodine atoms are added with the atoms from other release paths in Section 7.6.3A and then converted into the isotopic iodine activities in Table 10

CALCULATION NO. H-1-ZZ-MDC-1880 I REVISION NO. 4 1 PAGE NO. 18'of 117 using the atom/activity relation in Table 8. The iodine isotopic activities deposited on the CREF charcoal filter due to the containment, ESF, and MSIV leakages are shown in Table 13. 2.5.5.2 Post-LOCA Iodine and Aerosol Activity On CREF Charcoal/HEPA Filter - ESF Leakage Similarly, the iodine deposited on the CREF charcoal filter is calculated in Section 7.6.2 for the post-LOCA ESF leakage. The post-LOCA ESF leakage consists of a non-aerosol iodine release (97% of elemental iodine + 3% of organic iodine) (Ref. 10.1, Section 5.6) only, therefore, there is no aerosol mass deposited on the CR HEPA filter (HEPU300ES00.o0, CR Compartment Nuclide Inventory @ 720 hrs). The total (elemental + organic) iodine atoms deposited on the CREF charcoal filter due to the ESF leakage is calculated in Section 7.6.2. The iodine atoms are added with the atoms from other release paths in Section 7.6.3A and then converted into the isotopic iodine activities in Table 10 using the atom/activity relation established in Table 8. The iodine isotopic activities deposited on the CREF charcoal filter due to the containment, ESF, and MSIV leakages are shown in Table 13. 2.5.5.3 Post-LOCA Iodine and Aerosol Activity On CR Charcoal/HEPA Filter - MSIV Leakage The CREF intake filter iodine and aerosol activities are calculated in Section 7.6.3 for the MSIV leakage. The total aerosol mass deposited on the CREF HEPA filter due to the MSIV leakage is calculated, which is used with the aerosol mass/activity relation established in Table 11 to calculate the aerosol isotopic activities deposited on the CREF HEPA filter. This aerosol mass is added with the aerosol mass from the containment leakage in Section 7.6.3A and then converted into the isotopic aerosol activities in Table 12 using the atom/activity relation in Table 11. The total (elemental + organic) iodine atoms deposited on the CREF charcoal filter due to the MSIV leakage is calculated in Section 7.6.3. The iodine atoms are added with the atoms from other release paths in Section 7.6.3A and then converted into the isotopic iodine activities in Table 10 using the atom/activity relation established in Table 8. The iodine and aerosol isotopic activities deposited on the CREF charcoal/HEPA filter due to the containment, ESF, and MSIV leakages are shown in Table 13. 2.5.5.4 MicroShield Analysis of CR Charcoal/HEPA Filter Shine The total CREF charcoal/HEPA filter iodine and aerosol isotopic activities in Table 13 are input into the MicroShield (Ref. 10.9) Computer Run HCCRFLT2.MS5 with the source geometry, dimension, and detector location as shown in Figures 6 and 7 to compute the direct dose rate from the CREF filter. Due to the limitations of the MicroShield code, which calculates the dose rate at the dose point location within the projected area (width and height dimension), the dose point location at the center of the charcoal filter projected area is conservatively modeled at a distance of 20.0' from the ceiling because the actual dose point involves the slant distance of 30.99 feet (Section 7.6.4). The concrete thickness of 3.0' is conservatively modeled because the slant distance in the concrete shielding is 3.26' (Section 7.6.4). The 720-hrs direct dose from the CR filter shine is calculated in Section 7.6.5 using the CR occupancy factors and added to doses from other post-LOCA sources in Section 8.0. 2.6 CR and FRVS Vent Charcoal/HEPA Filter Efficiencies The CR and FRVS vent charcoal filters are tested to comply with Generic Letter 99-02 requirements (Refs. 10.3 and 10.6). However, since 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).

CALCULATION NO. H-1-ZZ-MDC-1880 I REVISION NO. 4 1 PAGE NO. 19 of 117 2.7 Determine Compliance of Increased Dose Consequences With 10CFR50.59 Guidance Consistent with the RG 1.183, Section 1.1.1, once the initial AST implementation has been approved by the staff and has become part of the facility design basis, the licensee may use 10 CFR 50.59 and its supporting guidance in assessing safety margins related to subsequent facility modifications and changes to procedures. The NRC Safety Evaluation Report for Amendments 134 and 146 (Refs. 10.49 and 10.50) approved the AST for the HCGS licensing basis analyses. An increase in control room, EAB or LPZ dose consequence is considered acceptable under the 10 CFR 50.59 rule if the magnitude of the increase is minimal (as defined by the guidance in Refs. 10.36 and 10.44), and if the total calculated dose is less than the allowable Regulatory Guide 1.183 dose limit. The current licensing basis analysis is documented in the calculation H-1-ZZ-MDC-1880, Rev 3. The increases in the expected EAB, LPZ, and CR doses are compared with the 10 CFR 50.59 allowable minimal dose increases in Section 8.2. Similarly, the proposed calculated total doses are compared with the allowable regulatory guide limits. The comparisons in Section 8.2 confirm that the expected increases in the EAB, LPZ, and CR doses and the total calculated doses are less than the corresponding minimal dose increases and allowable regulatory guide limits. Therefore, pursuant to 10 CFR 50.59 guidance as defined in References 10.36 and 10.44, the proposed increase in the core thermal power level and resulting post-LOCA doses can be adopted as current design and licensing bases for the HCGS. 3.0 ACCEPTANCE CRITERIA The following NRC regulatory requirement and guidance documents are applicable to this HCGS Alternative Source Term LOCA Calculation:

• Regulatory Guide 1.183 (Ref. 10.1)
• 10CFR50.67 (Ref. 10.4)
• Standard Review Plan section 15.0.1 (Ref. 10.18)

Dose Acceptance Criteria are: Regulatory Dose Limits Dose Type Control Room (rem) EAB and LPZ (rem) TEDE Dose 5 25

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO. 4 l PAGE NO. 20 of 117

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 3,917 MWt corresponding to current fuel enrichment and fuel burnup, which is 1.173 times the HCGS current licensed thermal power of 3,339 MW, (Ref. 10.6.9) including the 2% instrumentation uncertainty. Per Section 3.1 of RG 1.183, for DBA LOCA, all fuel assemblies in the core are assumed to be affected, therefore, the core average inventory is used in the analysis (Table ID). 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 I and 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 and Ref. 10.2 Section 2.2.2.1.2). 4.6.3 The HCGS drywell and suppression chamber may be purged for up to 500 hrs per year (Ref. 10.6.18). Normally, 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. Per

CALCULATION NO. H-1-ZZ-MDC-1880 I REVISION NO. 4 1 PAGE NO. 21 of 117 Reference 10.42, Safety Evaluation Section 3.1, the revised 10 CFR 50.44 no longer defines a design-basis LOCA hydrogen release, and eliminates requirements for hydrogen control systems to mitigate such a release. The installation of hydrogen recombiners and/or vent and purge systems required by 10 CFR 50.44(b)(3) was intended to address the limited quantity and rate of hydrogen generation that was postulated from a design-basis LOCA. The Commission has found that this hydrogen release is not risk-significant because the design-basis LOCA hydrogen release does not contribute to the conditional probability of a large release up to approximately 24 hours after the onset of core damage. In addition, these systems were ineffective at mitigating hydrogen releases from risk-significant beyond design-basis accidents (BDBA). Therefore, the Commission eliminated the hydrogen release associated with a design-basis LOCA from 10 CFR 50.44 and the associated requirements that necessitated the need for the hydrogen recombiners and the backup hvdrozen vent and purge systems. As a result, the HCGS deleted hydrogen recombiners from the its licensing basis (Ref. 10.6.2). 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. The calculation of these two components of the TEDE should consider all radionuclides, including progeny from the decay of parent 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 and 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 offsite dose analysis uses DDE in lieu of EDE Dose Conversion Factors in determining external exposure (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 and RGP 4.4, and Ref. 10.4). 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 and Ref. 10.4). LPZ Dose Acceptance Criteria: 25 Rem TEDE (50.67(b)(2)(ii))

CALCULATION NO. H-1-ZZ-MDC-1880 I REVISION NO. 4 [ PAGE NO. 22 of 117 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 and 4.4. These rates are incorporated in design inputs 5.7.2 and 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). 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 and 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 and 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

CALCULATION NO. H-1-ZZ-MDC-1880 I REVISION NO. 4 1PAGE NO. 23 of 117 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 inleakage is conservatively assumed to be 500 cfm (Design Input 5.6.7) during the CREF transition period of 30 minutes after a LOCA. This unfiltered inleakage includes 10 cfm for CR ingress and egress (Design Input 5.6.7). 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 ft 3 (3000 ft3/min x 2 min x 1.1 [for 10% variation in flow] = 6,600 ft3). The conservative assumption of 500 cfm unfiltered inleakage during the transition period would result in 15,000 ft 3 (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).

[CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO. 4 PAGE NO. 24 of 117 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). 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

CALCULATION NO. H-1-ZZ-MDC-1880 I REVISION NO. 4 1 PAGE NO. 25 of 117 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.

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO.4 PAGE NO.26 of 117 Design Input Parameter Value Assigned Reference 5.3 CONTAINMENT LEAKAGE MODEL PARAMETERS 5.3.1 Source Term 5.3.1.1 Thermal Power Level 3,917 MWt Section 7.9 5.3.1.2 Not Used 5.3.1.3 Isotopic Average Core Inventory (Ci/MWt) (Table ID) (Ref. 10.45, Appendix A) See Note Below Isotope Ci/MW, Isotope Ci/MWt Isotope Ci/MWt CO-58* 1.529E+02 RU103 7.700E+04 CS136 1.860E+03 CO-60* 1.830E+02 RU105 2.700E+04 CS137 6.760E+03 KR 85 3.330E+02 RU 106 2.940E+04 BA 139 4.950E+04 KR 85M 7.350E+03 RHI105 2.530E+04 BA 140 4.780E+04 RB 86 1.420E+04 SB127 2.800E+03 LA140 5.080E+04 KR 87 2.OOOE+04 SB 129 8.490E+03 LA141 4.5 1OE+04 KR 88 6.350E+01 TE127 2.780E+03 LA142 4.370E+04 SR 89 2.690E+04 TE127M 3.710E+02 CE141 4.540E+04 SR 90 2.640E+03 TE129 8.350E+03 CE143 4.220E+04 SR 91 5.300E+04 TE129M 1.240E+03 CE144 7.424E+04 SR 92 3.610E+04 TE131M 2.764E+04 PR143 4.080E+04 Y 90 2.810E+03 TE132 3.810E+04 ND147 1.810E+04 Y 91 3.440E+04 1131 2.670E+04 NP239 5.220E+05 Y 92 3.620E+04 1132 3.870E+04 PU238 9.040E+01 Y 93 4.160E+04 1133 5.510E-+04 PU239 1.090E+01 ZR 95 4.850E+04 1134 6.060E+04 PU240 1.410E+01 ZR 97 1.468E+05 1135 6.220E+04 PU241 4.090E+03 NB 95 4.870E+04 XE133 5.300E+04 AM241 4.600E+00 MO 99 5.100E+04 XE135 1.820E+04 CM242 1.090E+03 TC 99M 4.460E+04 CS 134 5.350E+03 CM244 5.240E+01

• 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 IC 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 5.3.1.5 Timing of Release Phase (Ref. 10.1, Table 4) Phase Onset Duration Gap Release 2 min 0.5 hr Gap release starts at 0.0 Early In-Vessel Release 0.5 hr 1.5 hr sec 5.3.1.6 Iodine Chemical Form Iodine Chemical Form  % Aerosol (CsI) 95.0% 10.1, RGP 3.5 Elemental 4.85% Organic 0.15%

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO. 4 PAGE NO. 27 of 117 Design Input Parameter Value Assigned Reference 5.3.1.7 Release Fraction (Ref 10.1, Table 1) BWR Core Inventory Fraction Released Into Containment Group Gap Release Phase Early In-Vessel Release Phase 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 5.3.1.8 Fuel Burnup 58 GWD/MTU < 62 GWD/MTU 10.45 and 10.1 5.3.1.9 Post-LOCA Drywell Temperature Post-LOCA Time (Hr) Temperature (OF) 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 5.3.2 Activity Transport in Primary Containment 5.3.2.1 Primary Containment Parameters 5.3.2.2 Drywell Air Volume 169,000 ft3 10.6.6 and 10.16 5.3.2.3 Suppression Chamber Air 137,000 ft3 10.6.6 and 10.16 Volume 5.3.2.4 Containment Air Volume 306,000 ft3 DI 5.3.2.2 + DI 5.3.2.3 5.3.2.5 Containment Leak Rate 0-720 hrs 0.5 v%/day 10.6.4 and 10.15 5.3.2.6 Draw Down Time 375 sec 10.6.8 5.3.2.7 Cont. Leakage Before Directly Released to Environment 10.1, RGP A.4.2 Draw Down Time (< 375 sec) 5.3.2.8 Cont. Leakage After Directly Released to Reactor 10.1, RGP A.4.2 Draw Down Time (>375 sec) Building 5.3.2.9 Reactor Building Volume 4,000,000 ft3 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 cfm +/- 10% 10.6.3, and 10.6.16 Rate Before Draw Down 5.3.2.12 FRVS Vent Exhaust 3324 + 5676e-1'- t Actual Eqn in Ref. 10.19, page 24 Flow Rate After Draw Down Flow Rates calculated in Design is 3324 + 5637e--18t Input 5.3.2.15

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO. 4 PAGE NO. 28 of 117 Design Input Parameter Value Assigned Reference 5.3.2.13 FRVS Vent Exhaust Filter Efficiency Iodine Species Efficiency (%) Elemental 90% 10.47 Aerosol 99% Section 7.7 Organic 90% 10.47 5.3.2.14 FRVS Recirc Filter Efficiency Iodine Species Efficiency (%) Elemental 0% Assumed Aerosol 99% Section 7.7 Organic 0% Assumed 5.3.2.15 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 + 5 6 7 6 e- I8t 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 5.3.2.16 FRVS Recirc Flow Rate 120,000 cfm - 10% 10.6.12 (or, 108,000 cfm) 5.4 ESF Leakage Model Parameters 5.4.1 Sump Water Volume 118,000 ft3 10.6.5 and 10.16 5.4.2 ESF Leakage 2.85 gpm Assumption 5.4.3 ESF Leakage Initiation 0 minute Assumption Time 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 and Tables 1 and 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 10% 10.1, RGP A.5.5, and 10.25, pages 35 through 45 5.4.7 Chemical Form Iodine In ESF Leakage Elemental 97% 10.1, RGP A.5.6 Organic 3% 5.4.8 Pool Peak Temperature 212.3'F 10.17, Attachment 1, PUSAR Table 1_ 1 4-1

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO. 4 PAGE NO. 29 of 117 Design Input Parameter Value Assigned Reference 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 100 scfh Assumed First Intact Line 5.5.4 Not Used 5.5.5 Number of Steam Lines 4 10.11 and 10.12e 5.5.6 Diameter and Wall Diameter = 26" 10.13b Thickness of Pipe Between RPV Wall Thickness = 1.117" 10.14c Nozzle and Inboard Isolation Valves HV F022A/B/C/D 5.5.7 Diameter and Wall Diameter = 26" 10.12e Thickness of Pipe Between Wall Thickness = 1.117" 10.14c Inboard and 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 and 3rd Isolation Valves HV 363 1A/B/C/D 5.5.9 Diameter of Pipe Between Diameter = 28" 10.12a 3rd Isolation and Turbine Stop Wall Thickness = 0.934" 10.14b Valves MSV1/2/3/4 5.5.10 Corrosion Allowance For 0.12" 10.14 Steam 5.5.11 Drywell Peak Pressure 50.6 psig 10.17, Section 3.3.1 5.5.12 Drywell Peak 2980 F 10.17, Section 3.3.1 Temperature 5.6 Control Room Model Parameters 5.6.1 CR Volume 85,000 ft3 10.33, Page 10 5.6.2 CREF System Flow Rate 1,000 cfm 10.6.16 1,050 cfm Assumed 5.6.3 CR Minimum Recirculation 2,600 cfm 10.6.15 Flow Rate 5.6.4 CR Unfiltered Inleakage 155 +/- 10 cfm actually measured 10.46, Table 19 After CREV System Initiation + 10 cfm for ingress and egress 10.23, Section 2.5 Total = 300 cfm Assumed 5.6.5 CREV System Initiation 30 minutes Assumption 4.8.5 Time After a LOCA 5.6.6 CR Charcoal and HEPA 99% Sections 7.7 and 7.9 Filter Efficiencies 5.6.7 CR Unfiltered Inleakage 10 cfm for ingress and egress 10.23, Section 2.5 Prior to CREV System Initiation 500 cfm (including 10 cfm for 10.40, page 6.4-8 and Assumption ingress and egress) 4.8.6

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO. 4 I PAGE NO. 30 of 117 Design Input Parameter Value Assigned Reference 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 and 10.29b 5.6.9 CR X/Qs For Containment and ESF Leakage Release Via FRVS Vent Ground Level Release Time X/Q (sec/m 3) 0-2 1.25E-03 10.5, page 34 2-8 8.09E-04 8-24 3.04E-04 24-96 2.1OE-04 96-720 1.59E-04 5.6.10 CR Occupancy Factors Time (Hr)  % 0-24 100 10.1, RGP 4.2.6 24-96 60 96-720 40 5.6.11 CR X/Qs For MSIV Leakage Release Via Turbine Building Louvers Ground Level Release Time X/Q (sec/m 3) 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 5.6.12 CR Breathing Rate 3.5E-04 (m 3/sec) 10.1, RGP 4.2.6 5.6.13 Minimum Reactor Bldg 1' -6" 10.35 Wall Thickness 5.7 Site Boundary Release Model Parameters 5.7.1 EAB X/Q (0-2 Hrs) 1.9E-04 sec/mr3 10.32, pages 5 and 9 5.7.2 EAB Breathing Rate 3.5E-04 m 3/sec 10.1 5.7.3 LPZ X/Qs (0-720 Hrs) Time X/Q (sec/m 3) 0-2 1.9E-05 10.32, pages 5 and 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

CALCULATION NO. H-l-ZZ-MDC-1880 REVISION NO. 4 PAGE NO. 31 of 117 Design Input Parameter Value Assigned Reference 5.7.4 Offsite Breathing Rates _I_ Time BR (m 3/sec) 0-8 3.5E-04 10.1, RGPs 4.1.3 and 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 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"

CALCULATION NO. H-1-ZZ-MDC-1880 I REVISION NO. 4 1 PAGE NO. 32 of 117 6.0 COMPUTER CODES AND COMPLIANCE WITH REGULATORY REQUIREMENTS 6.1 Computer Codes All computer codes used in this calculation have been approved for use with appropriate Verification and Validation (V & V) documentation. Computer codes used in this analysis include: RADTRAD 3.02 (Ref. 10.48): This is an NRC-sponsored code approved for use in determining control room and offsite doses from releases due to reactor accidents. This code was used by PSEG Nuclear in various AST license amendments, which are approved by the NRC. PSEG Nuclear performed in-house V&V of the code (Ref. 10.48). Therefore, the code is considered acceptable to be used for the HCGS AST analysis. RADTRAD 3.02 is used rather than the current RADTRAD 3.03 to maintain consistency with previous revisions of Calculation H-1-ZZ-MDC-1880. Per the RADTRAD 3.03 V&V report, the following technical modifications, which have minimal impact on the RADTRAD 3.02 dose results, have been incorporated into RADTRAD 3.03: Modifications to correct logic errors that existed in the previous version Multiple release paths from a compartment to the environment caused a significant conservative error in the control room dose, it became proportional to the number of paths Control room filter deposition used incorrect array (< 0.1% effect on calculated dose) Invalid filter loading values for all cases (no effect on calculated dose)

        -   Suppression pool decontamination used incorrect volume (NAI-1 1) (< 0.1% effect on calculated dose)
        -   A coefficient for the Gormley & Kennedy turbulent deposition model was in error (no effect on calculated dose)
        - Natural deposition model for APWR had a coefficient error (NAI- 12) (no effect on calculated dose)
        - Powers natural deposition model used a derived removal coefficient instead of the current value

(< 0.1% effect on calculated dose)

        - Dose conversion filename length could cause the code to terminate (no effect on calculated dose)
        - RADTRAD control of time steps to improve dose accuracy (RADTRAD v3.02) (< 1% effect on calculated dose)
        - Suppression pool decontamination that removed noble gases was corrected to allow their passage through the pool. (RADTRAD v3.02) (potential significant non-conservative effect on calculated dose)

Modification to the definition of a control room

        - This modification was essentially a change to the definition of a control room. The control room was defined to be a compartment not included in the mass balance. This allows the offsite dose to be independent of the existence of a control room. Previously, the offsite dose would change

(<1%) when a control room with a significant through flow was added. (NAI-7)

        - NRC Acceptance Test Case 16 (Table 8-1) originally called the auxiliary building a control room. As the mass balance excludes the control room, the input for this case was modified to allow a correct offsite and control room dose calculation. Doses can still be calculated in the auxiliary room by using an effective inlet X/Q and an iodine protection factor formulation as was done in the rebaselining (Callan 1998) or by executing the model twice, first with the control room modeled as the control room and second with the auxiliary building modeled as the control

CALCULATION NO. H-l-ZZ-MDC-1880 I REVISION NO. 4 1 PAGE NO. 33 of 117 room. This is the same procedure one would use to evaluate dose on the Technical Support Center. MicroShield 5.05 (Ref. 10.9): A commercially available and accepted code used to determine dose rates at various source-receptor combinations. Several runs were made at various times during the LOCA since the source strength varies over time. This code was used by PSEG Nuclear in various AST license amendments, which are approved by the NRC. PSEG Nuclear performed in-house V&V of the code (Ref. 10.9). Therefore, the code is considered acceptable to be used for the HCGS AST analysis.

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO. 4 PAGE NO. 34 of 117 7.0 CALCULATIONS 7.1 HCGS Plant Specific Nuclide Inventory File (NIF) For RADTRAD3.02 Input The RADTRAD nuclide inventory file HEPULOCA l_DEF.txt 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 core inventory information obtained from Reference 10.45. The RADTRAD NIF HEPULOCAIDEF.txt is modified based on the higher power core inventory in Table ID and used in the analyses. 7.2 Determination of MSIV Leak Rates 7.2.1 Analyzed Case The total leakage from all main steam lines is 250 scfh measured at 50.6 psig, allowing a maximum of 150 scfh from any one of the 4 main steam lines. The total containment leakage is 0.5 w%/day. No leakage reduction is credited after 24 hours. The total containment leakage does not include the leakage through the MSIVs. 7.2.2 MSIV Leakage During 0-2 hrs Note: The RADTRAD runs model MSIV leakage beginning at 0 minutes, which is prior to the 2 minute start of the gap release per Section 5.3.1.5. Total Drywell volume = 169,000 ft3 (Ref. 10.16) Total MSIV leakage measured @ 50.6 psig = 250 scfh Per the ideal gas law, PV= nRT or PV/T = nR. Given that nR is a constant for the air leakage, PV/T at post-LOCA conditions is equal to PV/T at STP conditions. P @LOCA = Drywell peak pressure = 50.6 psig (Ref. 10.17, Section 3.3.1) T @LOCA = Drywell peak temperature = 298°F (Ref. 10.17, Section 3.3.1) = 298'F + 460 = 758 0R P @STP = Standard pressure = 14.7 psia T @STP = Standard temperature = 68 0 F = 68°F + 460 = 528 0R V @STP = MSIV leakage based @ 50.6 psig = 250 scfh V @LOCA = (PV/T @STP) x (T/P @LOCA) 0-2 hrs MSIV leakage @ drywell peak pressure of 50.6 psig and temperature of 2980 F

     = 250 scfh x [14.7 psia / (50.6 psig + 14.7 psia)] x [758 0R / 528 0R]
     = 250 scfb x 0.225 x 1.436 = 80.78 cfh
     = (80.78 ft 3/hr x 24 hr/day) x 100% / 1.69E+05 ft3 =      1.147 %/day
     = (80.78 ft3 /hr) / (60 min/hr) = 1.346 cfm The 0-2 hrs 250 scfh MSIV leakage is released via two of the four Main Steam (MS) lines. A maximum allowable leak rate of 150 scfh is postulated from the shortest MS line with its inboard MSIV failed.

The remaining leak rate of 100 scfh is postulated from the shortest of the three intact MS lines (i.e., the second shortest of the four MS lines). No leakage is postulated from the remaining two intact MS Lines.

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO. 4 PAGE NO. 35 of 117 0-2 hrs allowable leakage from the MS line with a failed MSIV (at maximum 150 scfh leak rate)

       = (150 scfh / 250  scfb total) x 80.78 cfh = 48.47 cfh = 0.808 cfm 0-2 hrs allowable   leakage from the shortest intact MS line (at maximum 100 scfh leak rate)
       = (100 scfh / 250  scfh total) x 80.78 cfh = 32.31 cfh = 0.539 cfm 7.2.3   MSIV Leakage During 2-720 hrs Two hours after a LOCA the drywell and suppression chamber volumes are expected to reach an equilibrium condition and the post-LOCA activity is expected to be homogeneously distributed between these volumes. The homogeneous mixing in the primary containment will decrease the activity concentration and therefore decrease the activity release rate through the MSIVs. To model the effect of this mixing, the MSIV flow rate used in the RADTRAD model is decreased by calculating a new leak rate based on the combined volumes of the drywell and suppression chamber.

Drywell + Suppression Chamber free air volume = 306,000 ft3 (Design Input 5.3.2.4) 2-720 hrs MSIV leakage @ drywell peak pressure of 50.6 psig = 80.78 cfh (Section 7.2.2)

       = (80.78 cfh x 24 hr/day) x 100% / 3.06E+05 ft3 =            0.634   /dayI Corresponding MSIV leak rate = 80.78 cfh x (1.69E+05 ft3 / 3.06E+05 ft3) = 44.61 cfh 2-720 hrs allowable leakage from the MS Line with a failed MSIV (at maximum 150 scfh leak rate)
       = (150 scfh / 250 scfh total) x 44.61 cfh = 26.77 cfh = 0.446 cfm 2-720 hrs allowable leakage from the shortest intact MS Line (at maximum 100 scfh leak rate)
       = (100 scfh / 250 scfh total) x 44.61 cfh = 17.84 cfh = 0.297 cfm 7.2.4   Deleted.

7.2.5 MSIV Leakage To Environment: 7.2.5.1 MSIV Leakage into the pipe spool between outboard MSIV and TSV and environment from MSIV failed line (MSIV Failed MS Line 1) 0-720 hrs It is conservatively assumed that the MSIV leakage past the TSV expands to the atmospheric condition as follows: Upstream of inboard MSIV in MSIV failed line (Section 7.2.2): VI = 48.47 cfh P1 = 50.6 psig + 14.7 = 65.3 psia TI = (298°F + 460) = 758'R Downstream of outboard MSIV in MSIV failed line (Atmospheric Condition): V2 = TBD P2 = 14.7 psia T2 = (68°F + 460) = 528°R MSIV Leakage into the pipe spool between outboard MSIV and TSV and environment from MSIV failed line (MS Line 1): V2 = (PV/T @1) x (T/P @2) (65.3 psia x 48.47 cfh / 758 0 R) x (528 0 R / 14.7 psia) 150 cfh = 2.5 cfm

CALCULATION NO. H-1-ZZ-MDC-1880 I REVISION NO. 4 1 PAGE NO. 36 of 117 This is as expected, given that the 48.47 cfh leakage rate is equivalent to 150 scfh upstream of the outboard MSIV, and therefore it is equivalent to 150 cfh downstream of the outboard MSIV in the presence of standard pressure and temperature atmospheric conditions. 7.2.5.2 MSIV Leakage into the pipe spool between outboard MSIV and TSV and environment from MSIV shortest intact line (Intact MS Line 2) 0-720 hrs Upstream of inboard MSIV in the shortest intact MS Line (Section 7.2.2): VI = 32.31 cfh P1 = 50.6 psig+ 14.7 = 65.3 psia TI = (298°F + 460) = 7580 R Downstream of inboard MSIV in intact line (assumed Atmospheric Condition): V2 = TBD P2 = 14.7 psia T2 = (68°F + 460) = 528°R MSIV Leakage into the intact pipe spools between the inboard and outboard MSIVs and between the outboard MSIV and TSV (and environment) from the intact line: V2 = (PV/T @1) x (T/P @2)

       = (65.3 psia x 32.31 cfh / 758 0R) x (528 0R / 14.7 psia) 100 cfh = 1.667 cfm This is as expected, given that the pressure and temperature conditions in the intact pipe spools between the inboard and outboard MSIVs and between the outboard MSIV and TSV are assumed to be the same as the standard pressure and temperature atmospheric conditions present in the environment.

7.3 Main Steam Line Volumes and Surface Area for Plateout of Activity The two shortest main stream lines are selected for the aerosol deposition, namely the steam lines "A" and "D" connected to the reactor pressure vessel (RPV) nozzle N3A and N3D (Ref. 10.21). The steam line A is postulated to ruptured. The inboard MSIV connected to line A is postulated to fail to close and remains open during the accident to meet the single active failure requirement. The horizontal length of piping between the outboard MSIV and TSV becomes very critical in determining the aerosol deposition in the MSIV failed line, which contributes a major dose. The dimensions associated with the HCGS main steam piping in References 10.12, 10.13, and 10.21 are documented in the following section. 7.3.1 Piping Parameters: The piping parameters for the various piping segments between the RPV nozzle and TSV are listed in the following sections based on the pipe classes. The piping parameters are typical for all steam similar segments inside and outside drywell. 7.3.1.1 MSIV Line Between RPV Nozzle and Outboard Isolation Valve: Piping Class= DLA (Ref. 10.13b) Pipe Diameter = 26" (Ref. 10.13b) Minimum Wall Thickness = 1.117" (Ref. 10.14.c) Corrosion Allowance For Steam = 0.12" (Ref. 10.14c)

CALCULATION NO. H-1-ZZ-MDC-1880 I REVISION NO. 4 PAGE NO. 37 of 117 Total Minimum Thickness = 1.117" + 0.12" = 1.237" 26" Pipe ID = OD - (2 x Min Wall Thickness) = 26" - 2 x 1.237" 2

                                                                          = 23.526" 2   = 1.961' Pipe Flow Area = (7c/4) x (Pipe I.D.) = (3.14 / 4) x (1.961 ft) = 3.019 ft 2

Length of short radius (SR) elbow = RO Where R = radius of SR elbow = 26" and 0 = Angle subtended by elbow in radian = 90 degree/57.29 degree/radian Length of SR elbow = 26" x 90/57.29 = 40.84" = 3.4' (typical SR elbow length) SR Elbow Volume = 3.019 ft2 x 3.4 ft = 10.27 ft3 (typical 26" SR elbow volume) Length of long radius (LR) elbow = RO Where R = radius of LR elbow = 1. 5 x 26" = 39" = 3.25' and 0 = Angle subtended by elbow in radian = 90 degree/57.29 degree/radian Length of LR elbow 39" x 90/57.29 = 61.27" = 5.11' (typical LR elbow length) LR Elbow Volume = 3.019 ft2 x 5.11 ft = 15.43 ft3 (typical 26" LR elbow volume) 7.3.1.2 MSIV Line Between Outboard and Third Isolation Valve Piping Class = DBB (Ref. 10.11 and 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" 2

                                                                          = 23.714" 2
                                                                                      = 1.976' Pipe Flow Area = (7c/4) x (Pipe I.D.) = (3.14 / 4) x (1.976 ft) = 3.065 ft 2

7.3.1.3 MSIV Line Between Third Isolation Valve and Turbine Stop Valve Piping Class = DBC (Ref. 10.11 and 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' 2 2 Pipe Flow Area = (7/4) x (Pipe I.D.)2 = (3.14 / 4) x (2.158 ft) = 3.656 ft Length of short radius (SR) elbow = RO Where R = radius of elbow = 28" and 0 = Angle subtended by elbow in radian = 90 degree/57.29 degree/radian Length of SR elbow = 28" x 90/57.29 = 43.99" = 3.666' (typical SR elbow length) SR Elbow Volume = 3.656 ft2 x 3.666 ft = 13.40 ft3 (typical 28" SR elbow volume) 7.3.2. Piping Volume and Surface Area for Aerosol Deposition - MSIV Failed Line 7.3.2.1 Piping from Inboard MSIV V028 to Outboard MSIV V032 (Ref. 10.12.e and 10.21 .a): I Length of Pipe Between Inboard and Outboard Isolation Valves Distance between the RPV centerline and inner edge of inbloard I MSIV 21 '-8-5/8" + 2'-9-7/8" = 24'-6-1/2" = 24.542' (Ref. 10.21.a) Distance between outer edge of drywell main steam penetration and centerline of RPV

       = 49'-2-1/2" (Ref. 10.12.e)

CALCULATION NO. H-1-ZZ-MDC-1880 I REVISION NO. 4 1PAGE NO. 38 of 117 Distance between inner edge of inboard MSIV and outer edge of drywell main steam penetration

       = 49'-2-1/2' - 24'-6-1/2" = 24'-8" = 24.667' Distance between outer edges of drywell main steam penetration and outboard MSIV
       = 3'-10" + 5'-3" = 9'-1" Distance between inner edge of inboard MSIV and outer edge of outboard MSIV
       = 24'-8" + 9'-1" = 33.75' 2

Total Pipe Volume = Flow Area x Total Length V1 = 3.019 ft x 33.75 ft = 101.89 ft 3 = 2.89 m 3 Total projected horizontal pipe surface area = D x L (Horizontal Length)

       = 1.961 ft x 33.75 ft = 66.18 ft2 Total horizontal pipe surface area = 71 D x L
       = 3.14 x 1.961 ft x 33.75 ft = 207.82 ft2 =3 19.32 m2 Total horizontal pipe volume = 101.89 ft 7.3.2.2 Piping from Outboard MSIV V032 to Third MSIV V003 (Ref. 10.12.e):

Horizontal pipe = 7'- 11" - 26" = 5.75' Horizontal volume = Flow area x Length = 3.065 ft2 x 5.75 ft = 17.62 ft3 SR elbow volume = 10.27 ft3 (Section 7.3.1.1) Vertical pipe = 19'-6-1/2" -26" = 17.375' Vertical volume V = Flow area x Length = 3.065 ft2 X 17.375 ft = 53.25 ft3 SR elbow volume = 10.27 ft3 (Section 7.3.1.1) Horizontal pipe = 6'-0" + 5'-3" = 11.25' 3 2 Horizontal volume V = Flow area x Length = 3.065 ft x 11.25 ft = 34.48 ft 3 3 3 3 3 Total Pipe Volume = 17.62 ft3 + 10.27 ft + 53.25 ft + 10.27 ft + 34.48 ft = 125.89 ft Total projected horizontal pipe surface 2area = D x L (Horizontal Length)

       = 1.976 ft x (5.75' + 11.25') = 33.59 ft Total horizontal pipe surface area = 71 D x L
       = 3.14 x 1.976 ft x (5.75' + 11.25') = 105.48 ft2 = 9.80 m2 3           3          3 Total horizontal pipe volume = 17.62 ft + 34.48 ft = 52.10 ft 7.3.2.3 Piping from Third MSIV V003 to Turbine Stop Valve MSV 3 (Ref. 10.12.c):

Horizontal pipe = 38'-0" + 39'-0" + 41'-6-1/2" + 21'-6-1/2" - 28" = 137'-9"3 137.75' volume = Flow area x Length = 3.656 ft X 137.75 ft = 503.61 ft 2 Horizontal Horizontal SR elbow length = 3.666 ft (Section 7.3.1.3) Horizontal SR elbow volume = 13.40 ft3 (Section 7.3.1.3) Horizontal pipe = 14'-11" + 38'-0" + 38'-0" - 28" = 88'-7" = 88.583' Horizontal volume = Flow area x Length = 3.656 ft2 x 88.583 ft = 323.86 ft3 Horizontal SR elbow length = 3.666 ft (Section 7.3.1.3) Horizontal SR elbow volume = 13.40 ft3 (Section 7.3.1.3) Horizontal pipe = 11 '-1-1/2" + 32'-4" = 43'-5-1/2" = 43.458' 3 2 Horizontal volume = Flow area x Length = 3.656 ft x 43.458 ft = 158.88 ft Total pipe volume = 503.61 ft3 + 13.40 ft3 + 323.86 ft3 + 13.40 ft3 + 158.88 ft3 = 1,013.15 ft3 Total projected horizontal pipe surface area = D x L (Horizontal Length)

       = 2'158 ft x (137.75 ft + 3.666 ft + 88.583 ft + 3.666 ft + 43.458 ft) 2
       = 2.158 ft x 277.123 ft = 598.03 ft Total horizontal pipe surface area = 71 D x L

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO. 4 PAGE NO. 39 of 117

       = 3.14 x 2.158' x 277.123 ft = 1,877.81 ft23= 174.54 m2 Total horizontal pipe volume = 1,013.15 ft 7.3.2.4 Volume and Surface Area for Aerosol Deposition - Outboard MSIV to TSV MSV 3 Combined total pipe volume V 2 = Outboard MSIV to Third MSIV + Third 3MSIV to TSV            3 MSV 3 3
       = 125.89 ft3 (Section 7.3.2.2) + 1,013.15 ft (Section 7.3.2.3) = 1,139.04 ft = 32.28 m Combined total projected horizontal pipe surface area                     2 2
         -33.59 ft2_(Section 7.3.2.2) + 598.03 ft (Section 7.3.2.3) = 631.62 ft Combined total horizontal pipe surface area                             2 2
       = 9.8 m 2 (Section 7.3.2.2) + 174.54 m (Section 7.3.2.3) = 184.34 m Total horizontal pipe volume                                                  3
       = 52.10 ft3 (Section 7.3.2.2) + 1,013.15 ft (Section 7.3.2.3) = 1,065.25 ft 3

7.3.2.5 Volume and Surface Area for Aerosol Deposition - Inboard MSIV to TSV MSV 3 Combined total pipe volume V11 3 3 3

       =  101.89 ft3 (section 7.3.2.1) + 1,139.04 ft (section 7.3.2.4) = 1,240.93 ft = 35.17 m Combined total projected horizontal pipe surface area                  2 2
       = 66.18 ft2 (section 7.3.2.1) + 631.62 ft (section 7.3.2.4) = 697.8 ft Combined total horizontal pipe surface area
       = 207.82 ft2 (section 7.3.2.1) + 105.48 ft 2 (Section 7.3.2.2) + 1,877.81 ft 2 (Section 7.3.2.3) 2
       = 2,191.11 ft2 = 203.66 m Total horizontal pipe volume                                                  3 3
       = 101.89 ft3 (section 7.3.2.1) + 1,065.25 ft (section 7.3.2.4) = 1,167.14 ft 7.3.3. Piping Volume and Surface Area for Aerosol Deposition - Intact Steam Line (RPV Nozzle N3D) 7.3.3.1 Piping from RPV Nozzle N3D to Inboard MSIV V031 (Ref. 10.21 .b):

Nozzle elevation (Center Line) = 170' 1/2" Straight pipe = 17'-0-1/2" - 13'-0" = 4'-0-1/2" = 4.04' Subtraction for short radius elbow = 26" = 2'-2" = 2.17' Length of straight pipe = 4.04 - 2.17' 3= 1.87' Volume = 3.019 ft x 1.87 ft = 5.66 ft LR Elbow Volume = 15.43 ft 3 (Section 7.3.1.1) SR elbow volume = 10.27 ft3 (Section 7.3.1.1) Height of 10'-10" Bend = 11'-7/14" / Sin 750 = 12'-0" Elevation of bend = 123'-6" + 12'-0" = 135'-3" Vertical Pipe = 170'-1-1/2" - 135'- 3" - 2'2" = 32'- 8-1/2" 3 32.708' 2 Vertical Volume, V = 3.019 ft x 32.708 ft = 98.75 ft 3 SR elbow V = 10.27 ft Length of bend = 11 '-7-1/4" - 5.11 = 6.494' Vertical Volume, V = 3.019 ft2 x 6.494' = 19.61 ft3 3 LR elbow V = 15.43 ft Horizontal bend pipe = 4'-4-15/16 + 9'-2-3/8" = 13.609' Volume = 3.019 ft x 13.609 ft = 41.09ft 3 3 LR elbow V = 15.43 ft Vertical pipe = 123'-5-1/2" - 107'-0-3/8"- 39" = 13.18' Vertical volume = 3.019 ft2 x 13.18 ft = 39.79 ft3

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO. 4 PAGE NO. 40 of 117 3 LR elbow V = 15.43 ft Horizontal length = 4'-7" - 39" = 1.33' 3 2 Horizontal volume = 3.019 ft x 1.33 ft = 4.02 ft 3 Horizontal LR elbow = 3.019 ft x 5.11 ft = 15.43 ft 2 Horizontal length = 3'-6-1/4" = 3.521' 3 2 Horizontal volume = 3.019 ft x 3.521 ft = 10.63 ft Total Volume of Steam Header From RPV Nozzle A To Inboard MSIV AO-80A 2VATI3 3 3 3 +41.09 ft 3 + 15.43 ft + 39.79 ft 3 +

       = 5.66 ft3 + 10.27 ft3 + 98.75 ft3 + 10.27 ft + 19.61 ft + 15.43 ft  3 3               3 15.43 ft3 + 4.02 ft3 + 15.43 ft + 10.63 ft = 301.81 ft = 8.55 m 3

Total horizontal pipe length = 1.87' + 13.609' + 1.33' + 5.11' + 3.521' = 25.44' Total projected horizontal pipe surface area = D x L = 1.961 ft x 25.44 ft = 49.89 ft22 Total horizontal pipe surface area = 7c D x L = 3.14 x 1.961 ft x 25.44 ft = 156.65 ft3 3 3 3 3 Total horizontal pipe volume = 5.66 ft3 + 41.09 ft + 4.02 ft + 15.43 ft + 10.63 ft = 76.83 ft 7.3.3.2 Piping from Inboard MSIV V031 to Outboard MSIV V035 (Ref. 10.12.e and 10.21.b): Same as in Section 7.3.2.1 3 3 Pipe Volume = Flow Area x Total Length = 101.89 ft = 2.89 m Total Total projected horizontal pipe surface area = D x L (Horizontal Length) = 66.18 ft2 Total horizontal pipe surface area = t D x3 L = 3.14 x 66.18 ft2 = 207.81 ft2 = 19.32 m2 Total horizontal pipe volume = 101.89 ft 7.3.3.3 Piping from Outboard MSIV 035 to Third MSIV V006 (Ref. 10.12.e): Same as in Section 7.3.2.2 3 Total Pipe Volume = 125.89 ft 2 Total projected horizontal pipe surface area = D x L (Horizontal Length) = 33.59 ft 2 2 Total projected horizontal pipe surface area = 71D x L = 3.14 x 33.59 ft = 105.47 ft = 9.80 m2 3 Total horizontal pipe volume = 52.10 ft 7.3.3.4 Piping from Third MSIV V006 to Turbine Stop Valve MSV 2 (Ref. 10.12.b): Horizontal pipe = 38'-0" + 39'-0" + 41'-6-1/2" + 15-7-1/2" - 28" 131 '-10"3 = 131.833' 2 Horizontal volume = Flow area x Length = 3.656 ft x 131.833 ft = 481.98 ft Horizontal SR elbow length = 3.666 ft (Section 7.3.1.3) Horizontal SR elbow volume = 13.40 ft3 (Section 7.3.1.3) Horizontal pipe = 6'7" + 40'-0" + 43'6" - 28" = 87'-9" = 87.75' 3 volume = Flow area x Length = 3.656 ft X 87.75 ft = 320.81 ft 2 Horizontal 3 Horizontal SR elbow = 13.40 ft Horizontal pipe = 26'-4-1/2" + 23'-0" = 49'-4-1/2" = 49.375' Horizontal 2 volume Flow area x Length = 3.656 ft x 49.375 ft = 180.52 ft3 3 3 3 3 3 Total pipe volume 481.98 ft3 + 13.40 ft + 320.81 ft + 13.40 ft + 180.52 ft = 1,010.11 ft Total projected horizontal pipe surface area = D x L (Horizontal Length)

       = 2.158 ft x (131.833 ft + 3.666 ft  2
                                              + 87.75 ft + 3.666 ft + 49.375 ft)
       = 2.158  ft x 276.29 ft =  596.23 ft

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO. 4 PAGE NO. 41 of 117 Total horizontal pipe surface area = 7 D x L

       = 3.14 x 596.23   ft2 = 1,872.16 ft2 = 174.02 3m2 Total horizontal  pipe volume = 1,010.11 ft 7.3.3.5 Volume and Surface Area for Aerosol Deposition - RPV to Outboard MSIV 035:

Combined total pipe volume V 3 = RPV Nozzle to Inboard MSIV + Inboard 3 MSIV to3 Outboard MSIV

       = 301.81 ft3 (Section 7.3.3.1) + 101.89 ft (Section 7.3.3.2) = 403.70 ft = 11.44 m 3

Combined projected total horizontal pi e surface area

       = 49.89 ft2 (Section 7.3.3.1) + 66.18 ft (Section 7.3.3.2) = 116.07 ft2 Combined total horizontal pipe surface area                                2 2
       = 156.65 ft2 (Section 7.3.3.1) + 207.81 ft (section 7.3.3.2) = 364.46 ft Combined total horizontal pipe volume                                     3 3
       = 76.83 ft3 (Section 7.3.3.1) + 101.89 ft (Section 7.3.3.2) = 178.72 ft 7.3.3.6 Volume and Surface Area for Aerosol Deposition - Outboard MSIV 035 to TSV MSV 3 Combined total pipe volume V4 = Outboard MSIV to Third MSIV + Third 3MSIV to TSV             3 MSV 3
       = 125.89 ft3 (Section 7.3.3.3) + 1,010.11 ft (Section 7.3.3.4) = 1,136.00 ft = 32.19 m 3

Combined projected total horizontal pipe surface area 2

       = 33.59 ft _(Section 7.3.3.3) + 596.23 ft (Section 7.3.3.4) = 629.82 2                                                            ft2 Combined total horizontal pipe surface area                                      2 2
       = 105.47 ft2 (Section 7.3.3.3) + 1,872.16 ft (Section 7.3.3.4) =1,977.63 ft Combined total horizontal pipe volume                                          3 3
       = 52.10 ft3 (Section 7.3.3.3) + 1,010.11 ft (Section 7.3.3.4) = 1,062.21 ft 7.3.3.7 Volume and Surface Area for Aerosol Deposition - RPV to TSV MSV 3 Combined total pipe volume V 33                                                3           3
       = 403.70 ft3 (section 7.3.3.5) + 1,136.00 ft (section 7.3.3.6) = 1,539.70 ft = 43.63 m 3

Combined total projected horizontal pipe surface area 2 2

       = 116.07 ft2 (section 7.3.3.5) + 629.82 ft (section 7.3.3.6) = 745.89 ft Combined total horizontal pipe surface area                                                  2
       = 364.46 ft2 (section 7.3.3.5) + 1,977.63 ft (section 7.3.3.6) = 2,342.09 ft2 = 217.69 m 2

Combined total horizontal pipe volume

       = 178.72 ft3 (section 7.3.3.5) + 1,062.21 ft 3 (section 7.3.3.6) = 1,240.93 ft3 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. RG 1.183, Appendix A, Section 6.5 requires that the components and piping systems used in the release path are capable of performing their safety function during and following a SSE. The main steam lines credited in the MSIV

CALCULATION NO. H-1-ZZ-MDC-1880 I REVISION NO. 4 1 PAGE NO. 42 of 117 leakage path are qualified to withstand the SSE, therefore, these lines are credited for the aerosol deposition in the following section: 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). The following aerosol deposition model is used, which is accepted by the Staff in Reference 10.22, Appendix A). 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 4 0 th 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/sec) Constant (hr 1) 60th (average) 0.00148 11.43 50th (median) 0.00117 9.04 40th 0.00081 6.26 10oth 0.00021 1.62 The Staff concluded that use of a 10 th percentile settling velocity with a well-mixed model is overly conservative and not appropriate (Ref. 10.22, page 11). Instead, the Staff believes it is acceptable to utilize median values (i.e., 50th percentile settling velocity) (Ref. 10.22, page 11). This analysis is conservative relative to the Staff s recommendation, in that it models a 4 0 th percentile settling velocity for aerosol deposition in the MSIV leakage paths beyond the inboard MSIVs. The derivation of the staff s well-mixed model begins with a mass balance as follows (Ref. 10.22, Page A-2): V

• dC=Q
• Cin,-Q
• C-ks *V
• C (1) dt Where V = volume of well-mixed region C = concentration of nuclides in volume Q = volumetric flow rate into volume ks = rate constant for settling And k,= u
• A V

Where u, = settling velocity A = settling area The aerosol settling velocities in the different control volumes are calculated in Table 5 using the above

    'equation based on the horizontal pipe projected areas and well mixed horizontal volumes obtained from Tables 2 through 4 and Sections 7.3.

Under steady-state condition, the derivative in the above equation (1) becomes zero. Equation (1) can be simplified as follows:

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO. 4 PAGE NO.43 of 117 C =_Cin I 1+ _**V Q RADTRAD allows input of filter efficiency for each flow path. Noting that C is also the concentration of nuclides leaving the volume, the above equation can be used to determine an equivalent filter efficiency as follows: lfiltl- C =1- 1 (2) Cin 1+ 2_.*V Q The aerosol removal efficiencies are calculated in Table 6 using Equation (2). The rate constant for settling velocity X, is calculated in Table 5 using the horizontal settling surface area and volume from Table 4. The aerosol removal efficiencies are calculated in Table 6 using the well-mixed pipe volumes from Table 3 and the volumetric full flow rates of 150 and 100 scfh in the MSIV failed and intact steam lines, respectively. 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 6f iodine on the pipe surface. Reference 10.52, which is Reference A-9 of Regulatory Guide 1.183 is used to determine the deposition and resuspension rates of elemental iodine as follows: di = elemental iodine vapor deposition velocity (cm/s)

     =  e( 280 9/T - 12.80 (+/- 0.33)) =  e(2 809/T - 12.5) (Ref. 10.52, pages 4 and 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 IF based the post-LOCA drywell temperature shown in Design Input 5.3.1.9. The elemental iodine deposition rate Xed (hr 1 ) = di

• S *3600 (Ref. 10.52, page 4)

V Where di = deposition velocity (m/sec) S = surface area of deposition (mi2 ) V = volume (mi3 ) The deposition velocity in cm/sec (which is converted into m/sec) and elemental iodine deposition rates at various drywell temperatures are calculated in Tables 1G and 1H for the MSIV failed line Volume VII and Intact line Volume V33, 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.

CALCULATION NO. H-l-ZZ-MDC-1880 I REVISION NO. 4 PAGE NO. 44 of 117 Resuspension rate of elemental iodine (sec- 1)

    = 2.32 (+/-2.00) x 10"'e-6°°' = 4.32 x 10"'e&600/7 Resuspension rate of elemental iodine ,er (hr-1) 00
    =  4.32 x 3600 x 10- e-6 /T The resuspension rates of elemental iodine at various drywell temperatures are calculated in Table 11.

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 IJ and 1K for the MSIV failed line Volume VII and intact line Volume V33, respectively. Net Deposition Rate of Elemental Iodine ke = X*ed - Xr I/DF = 1 - il = exp(-ke*t) (Ref 10.2, Equations 4 and 5, page 196) Where DF = decontamination factor rl = filter efficiency for elemental iodine ke = elemental iodine removal rate (hr-1) t = time (hr) Therefore, Elemental Iodine Filter Efficiency = 1 - e-(Ie

• t)

The net elemental iodine deposition rates (Xe) are obtained from Tables IJ and 1K and the corresponding filter efficiencies at various drywell temperatures are calculated in Tables IL and IM for the MSIV failed line Volume VII and intact line Volume V33, 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 2.85 gpm, which is doubled and converted into cfm as follows: 2.85 gallon/min x 2 x 1/7.481 ft3/gallon = 0.762 cfm 10% of ESF leakage becomes airborne = 0.1 x 0.762 cfn = 7.62E-02 cfm 7.6 Post-LOCA CREF Filter Shine Dose The post-LOCA CREF filter shine doses due to the containment, ESF, and MSIV leakages are calculated in the following sections

CALCULATION NO. H-l-ZZ-MDC-1880 I REVISION NO. 4 IPAGE NO. 45 of 117 7.6.1 Iodine/Aerosol Deposition on CREF Charcoal/HEPA Filter - Containment Leakage: Iodine Activity Deposited On CREF Charcoal Filter: As shown in Figure 5, the CR intake and recirculation charcoal filter elemental and iodine removal efficiency is 99% with the intake and recirculation flow rates of 1,050 cfm and 2,600 cfm, respectively. Charcoal Filter 71= 99% 'in 10 1050 'in 10.5 1039.5 'in Suppose Air Intake Iodine Activity (atoms) = 'in and Iodine Activity Deposited on Recirc Filter = Ide Activity deposited on the intake charcoal filter = in x 0.99 x 1,050 cfm = 1039.5 'in Filtered inflow activity introduced into the CR = 'in x (1 - 0.99) x 1,050 cfm = 10.5 Iin 10.5 'in + 300 'in (unfiltered inleakage) = 310.5 'in Recirc Charcoal Filter 1 = 99% 3.1 lin Ide 307.4 'in Activity deposited on recirc charcoal filter (conservatively assuming only one pass through the recirc charcoal filter) = Ide = 0.99 x (310.5 Iin cfm) = 307.4 lin ; Rearranging: lin = Ide / 307.4 Therefore, the activity deposited on the intake charcoal filter = 1039.5 'in = (1039.5/307.4) Ide Total elemental and organic iodine activity deposited on intake + recirc charcoal filter

     = (1039.5/307.4 Ide) + Ide = 4.38 Ide Ide = 1.1742E+14 Atoms (Elemental Iodine) + 4.6921E+14 Atoms (Organic Iodine) (HEPU300CLOO.oO, CR Recirculating Filter Nuclide Inventory @ 720 hrs)
     = 5.8663E+14 Atoms Total iodine activity deposited on the CR intake + recirc charcoal filter due to containment leakage
     = 4.38 x 5.8663E+14 Atoms = 2.5694E+15 Atoms Aerosol Mass Deposited On CREF HEPA Filter:

As shown in Figure 5, the CR intake and recirculation HEPA filter aerosol removal efficiency is 99% for the intake and recirculation flow rates of 1,050 cfm and 2,600 cfm respectively.

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO. 4 PAGE NO. 46 of 117 II Intake HEPA Filter I 1050 Ain ] 71= 99% 10.5 Ain 1039.5 Ain Suppose aerosol mass in intake air = Ain and aerosol mass deposited on recirc filter = Ade Aerosol mass deposited on HEPA filter = Ain x 0.99 x 1,050 cfm = 1039.5 Ain Filtered inflow aerosol mass introduced into the CR = Ain x (1 - 0.99) x 1,050 cfm = 10.5 Ain 10.5 Ain + 300 Ain (unfiltered inleakage) =310.5 Ain II Recirc HEPA Filter 3.1 Ain r!1

                                                     = 99%

Ade 307.4 Ain Aerosol mass deposited on recirc HEPA filter (conservatively assuming only one pass through the recirc HEPA filter) = Ade = 0.99 x (310.5 Ain cfm) = 307.4 Ain; Rearranging: Ain = Ade / 307.4 Therefore, the aerosol mass deposited on intake HEPA filter = 1039.5 Ain= (1039.5/307.4) Ade Total aerosol mass deposited on intake + recirc HEPA filter

     = (1039.5/307.4 Ade) + Ade = 4.38 Ade Ade  = 8.4925E-09 kg (HEPU300CLOO.oO, CR Recirculating Filter Nuclide Inventory @ 720 hrs)

Total aerosol mass deposited on CR intake + recirc HEPA filter due to containment leakage

     = 4.38 x 8.4925E-09 kg = 3.7197E-08 kg 7.6.2 Iodine/Aerosol Deposition on CREF Charcoal/HEPA Filter - ESF Leakage:

Iodine Activity Deposited On CREF Charcoal Filter: As discussed in Section 7.6.1 above: Total elemental and organic iodine activity (atoms) deposited on intake + recirc charcoal filter = 4.38 Id, Ide = 6.4972E+16 Atoms (Elemental Iodine) + 2.0094E+15 Atoms (Organic Iodine) (HEPU300ES00.o0, CR Recirculating Filter Nuclide Inventory @ 720 hrs)

     = 6.6981E+16 Atoms.

Total iodine atoms deposited on the CREF intake + recirc charcoal filter due to the ESF leakage

     = 4.38 x 6.6981E+16Atoms = 2.9338E+17 Atoms Aerosol Mass Deposited On CREF HEPA Filter:

Since post-LOCA ESF leakage consists of only a non-aerosol iodine release (97% of elemental iodine + 3% of organic iodine) (Ref. 10.1, Section 5.6), there is no aerosol deposited on the CREF intake + recirc HEPA filter (HEPU300ES00.o0, CR Compartment Nuclide Inventory @ 720 hrs). 7.6.3 Iodine/Aerosol Deposition on CREF Charcoal/HEPA Filter - MSIV Leakage:

CALCULATION NO. H-1-ZZMDC-1880 REVISION NO. 4 PAGE NO.47 of 117 Iodine Activity Deposited On CREF Charcoal Filter: As discussed in Section 7.6.1 above: Total elemental and organic iodine activity (atoms) deposited on intake + recirc charcoal filter = 4.38 Ide Ide = 5.7277E+14 Atoms (Elemental Iodine) + 2.8004E+15 Atoms (Organic Iodine) (H1N300MS00.o0, CR Recirculating Filter Nuclide Inventory @ 720 hrs)

      = 3.3732E+15 Atoms.

Total iodine atoms deposited on the CREF intake + recirc charcoal filter due to the MSIV leakage

      = 4.38 x 3.3732E+15 Atoms = 1.4775E+16 Atoms Aerosol Mass Deposited On CREF HEPA Filter:

As discussed in Section 7.6.1 above: Total aerosol mass deposited on intake + recirc HEPA filter = 4.38 Ade Ade = 2.2329E-08 kg (H1N300MS00.o0, CR Recirculating Filter Nuclide Inventory @ 720 hrs) Total aerosol mass deposited on CREF intake + recirc HEPA filter due to the MSIV leakage

      = 4.38 x 2.2329E-08 kg = 9.7801E-08 kg 7.6.3A Total Iodine/Aerosol Deposition on CREF Charcoal/HEPA Filter Total iodine atoms deposited on the CREF charcoal
      = Iodine atoms from containment leakage + Iodine atoms from ESF leakage + Iodine atoms from MSIV leakage
      = 2.5694E+15 Atoms + 2.9338E+17 Atoms + 1.4775E+16 Atoms = 3.1072E+17 Atoms, which is used in Table 10 to obtain the total iodine isotopic activities on the CREF charcoal.

Total aerosol mass deposited on the CREF H1EPA Filter

      = Aerosol mass from containment leakage + Aerosol mass from ESF leakage + Aerosol mass from MSIV leakage
      = 3.7197E-08 kg + 0.00 kg + 9.7801E-08 kg = 1.3500E-07 kg, which is used in Table 12 to obtain the total aerosol isotopic activities on the CREF HEPA Filter.

7.6.4 CREF Shielding Model The CREF unit location with respect to the center line of CR console is measured from the full scale drawings and used to determine the slant distance and angle through the concrete ceiling as follows:

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO. 4 PAGE NO. 48 of 117 Elevation View Plan View Horizontal distance - [(27'-2",)2 + (4'-4-1/2")2] 1/2 = 27.52' CREF center line elevation

     = CR ceiling elevation + Thickness of concrete base + V2 CREF height [see Figure 6]
     = 155'-3" (Ref. 10.39.c) + 6" (Ref. 10.39.c) + / 2 (3'-0") =157'-3Y Vertical distance between CR operator and CREF unit
     = 157'-3" - (137'-0" + 6'-0") = 157'-3" - 143'-0" = 14'-3" Assuming a 6 foot tall CR operator standing on the CR floor @ 137'-0".

Slant Distance = [(14.25')2 + (27.52 )2] 1/2 = 30.99' Slant Angle Through Concrete Ceiling = 8 = tan-' (14.25/27.52) = 27.380 Slant shielding thickness for 1'-6" concrete ceiling over the CR

     = 1.5'/sin 27.380 = 1.5'/0.46 = 3.26' The CR operator dose from the CREF shine dose is conservatively calculated using the concrete shielding of 3.0' and distance of 20' without taking the credit of shadow shielding of structure steel and equipment and CREF housing steel.

7.6.5 CR Direct Dose From Filter Shine CR Filter Shine Dose Rate = 4.065E-02 mRem/hr (MicroShield Run HCCRFLT2.MS5) CR Operator Exposure Time = I 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
     = 4.065E-02 mRem/hr x 1/1000 Rem/mRem x 316.8 hr = 1.29E-02 Rem 7.7   FRVS Vent and Recirc. and CR Charcoal/HEPA Filters Efficiencies HEPA Filter:

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

CALCULATION NO. H-1-ZZ-MDC-1880 I REVISION NO. 4 IPAGE NO. 49 of 117 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% - rq)/safety factor = (100% - i)/2 Where ri = HEPA filter efficiency to be credited in the analysis 0.05% = (100% - il)/ 2 0.1%= (100%-il) I = 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% - il)/safety factor = (100% - il)/ 2 Where il = CREF charcoal filter efficiency to be credited in the analysis CFREF Charcoal Filter 0.5%= (100%-iq)/ 2 1%= (100%-1l) fl = 100%- 1%= 99% FRVS Vent Charcoal Filter Elemental Iodine il = 90% (Ref. 10.47) Organic Iodine il = 90% (Ref. 10.47) Safety Grade Filter Efficiency Credited (%) Filter Aerosol Elemental Organic 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 14 through 18. These isotopic activities are obtained from the RADTRAD computer Runs HEPU300CLOO.oO, HEPU300ESOO.oO, and HIN300MS00.o0. This information is used in the MIDAS computer code to assess dose profile during a design basis accident for Emergency Planning. 7.9 Expected Higher Core Power Level Current Licensed Power Level = 3,339 MWt (Ref. 10.34) Proposed Power Level Increase = 15% Instrument Uncertainty = 2% (Ref. 10.10) Expected Higher Core Power Level = 3,339 MWt x 1.15 x 1.02

• 3,917 MWt 7.10 Drywell Wetted Surface Area The drywell surface area is calculated in RefereInce 10.24. The use of a smaller wetted surface is conservative because it results in a smaller elemnental removal coefficient and it takes a longer time to reach an elemental iodine decontamination factor (DF) of 200, which allows the elemental iodine to

CALCULATION NO. H-1-ZZ-MDC-1880 I REVISION NO. 4 1 PAGE NO. 50 of 117 remain airborne in the drywell atmosphere for release to the atmosphere via containment and MSIV leakage. Total drywell surface area = 152,261 ft2 excluding the RPV surface area (Ref. 10.24, page 15) The surface areas below the drywell spray ring are subject to be wetted in the spray solution, which are listed as follows based on areas presented in Ref. 10.24, page 15. The drywell wetted surface area is conservatively minimized as discussed in Section 2.1.3: Estimated 25% of drywell lining surface = 4,463 ft 22 (17,850 ft2 /4 = 4,463 ft2) Downcomer (to water level) = 3,168 ft2 Vent header and line = 9,727 ft Suppression chamber (to water level) = 15,408 ft2 Estimated 50% of major equipment = 5,306 ft2 (10,612 ft2 /2 = 5,306 ft2) Estimated 50% of structures = 6,181 ft 2 (12,361 ft 2 /2 =6,181 ft2) Total estimated wetted drywell surface area = 44,253 ft2 (69,126 ft 2 without reductions) 75% of total estimated wetted drywell surface area 33,200 ft2 (0.75 x 44,253 2 ft ) 7.11 Containment Elemental Iodine Removal Coefficient Natural deposition on containment surfaces (plateout) of the elemental iodine released to containment is calculated using the methodology outlined in NUREG-0800, Standard Review Plan 6.5.2 (Ref. 10.41, page 6.5.2-10) as follows: The equation for the elemental iodine removal by adsorption on wetted surface area is: X, = Kw x A/V Where: w= first order removal coefficient by wall deposition Kw mass transfer coefficient = 4.9 m/hr (Ref. 10.41, page 6.5.2-10) A = wetted surface area = 33,200 ft2 (Section 7.7) V = drywell net free air volume = 1.69E+05 ft3 for < 2.0 hrs and 3.06E+05 ft 3 for > 2.0 hrs Elemental iodine removal coefficient for < 2.0 hrs k, = Kw x A/V = 4.9 m/hr x (3.2808 ft/m) (33,200 ft2 ) / (1.69E+05 ft3) = 3.16 hr-1 Elemental iodine removal coefficient for > 2.0 hrs kw = K, x A/V = 4.9 m/hr x (3.2808 ft/m) (33,200 ft2) / (3.06E+05 ft3) = 1.74 hr-' Maximum DF of elemental iodine = 200 The containment leakage case is analyzed in RADTRAD Run CUTOFF.oO using the above calculated elemental iodine removal coefficients along with the information in Table 9 to determine the cutoff time for terminating elemental iodine removal from the containment atmosphere, which is 4.03 hrs (CUTOFF.oO). The cutoff time of 4.0 hrs is used in the containment and MSIV leakage path releases to the atmosphere.

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO. 4 PAGE NO. 51 of 117 8.0 RESULTS

SUMMARY

The results of AST analyses are summarized in the following sections: 8.1 The post-LOCA EAB, LPZ, and CR doses are summarized in the following table: Post-LOCA Post-LOCA TEDE Dose (Rem) Activity Release Receptor Location Path Control Room EAB LPZ Containment Leakage 5.37E-01 3.87E-01 1.47E-01 (3.1 hr) ESF Leakage 2.88E+00 5.22E-01 2.64E-01 (14.2 hr) MSIV Leakage 6.13E-01 8.68E-01 2.63E-01 (3.0 hr) Containment Purge O.OOE+00 0.OOE+00 O.OOE+00 Containment Shine O.OOE+00 0.OOE+00 O.OOE+00 External Cloud 0.OOE+00 0.OOE+00 0.OOE+00 CR Filter Shine 1.29E-02 O.OOE+00 0.OOE+00 Total 4.04E+00 1.78E+00 6.74E-01 Allowable TEDE Limit 5.0 E+00 2.50E+01 2.50E+01 RADTRAD Computer Run No. Containment Leakage HEPU300CLOO.oO HEPU300CLOO.oO HEPU300CLOO.oO ESF Leakage HEPU300ESOO.oO HEPU300ESOO.oO HEPU300ESOO.oO MSIV Leakage H1N300MSOO.oO HIN300MSOO.oO H1N300MSOO.oO

CALCULATION NO. H-1-ZZ-MDC-1880 I REVISION NO. 4 ] PAGE NO. 52 of 117 8.2 COMPLIANCE OF REVISED DOSES WITH THE 10 CFR 50.59 RULE IS SHOWN IN THE FOLLOWING TABLE: Current Proposed Regulatory Proposed Minimal SRP Design Basis Accident Total Total Dose Dose Dose Increase Dose Dose Dose Limit Increase Per 10CFR50.59 Limit (rem) (rem) (rem) (rem) (rem) (rem) TEDE TEDE TEDE TEDE TEDE TEDE A B C D=B-A E=O.1(C-A) F Loss of Coolant Accident H-1-ZZ-MDC- H-1-ZZ-MDC-1880, Rev 4 (LOCA) 1880, Rev33-M 18,e Control Room 4.17 4.04 5 -0.13 0.083 5 Exclusion Area Boundary 1.43 1.78 25 0.35 2.35 25 Low Population Zone 0.548 0.674 25 0.13 2.44 25 C From 10 CFR 50.67 (Ref. 10.4) F From Standard Review Plan 15.0.1, Section I1 (Ref. 10.18) 8.3 Dose Impact Due to Changes in Input Parameters and Methodology in Revisions 2, 3, and 4 This calculation determines control room, EAB, and LPZ doses due to post LOCA radioactivity releases from containment via three release pathways, i.e., containment leakage, ESF leakage, and MSIV leakage. Changes in input parameters and methodology that affect the quantity of radioactivity released via a pathway or that affect the quantity of radioactivity that reaches a dose location have an impact on calculated doses. Changes incorporated into Revisions 3 and 4 of this calculation that are expected to have significant dose impacts are:

1) the increase in ESF leakage from I gpm to 2.85 gpm (Table lE, Item IV-I),

2) the change in MSIV leakage volumetric flow rate after converting the volumetric flow rate from measured values at standard temperature and pressure conditions to drywell post LOCA atmospheric conditions (Table IE, Item V-3),

3) changing the MSIV release pathway along the failed and intact main steam lines from two well mixed volumes from both the failed and intact main steam lines to one well mixed volume for both the failed and intact main steam lines (Table 1E, Items V-4 and V-5),

4) crediting removal of elemental iodine via wall deposition on wetted containment surfaces (Table 1E, Item 1-2),

5) changes to containment air dilution volume, i.e., changing from instantaneous mixing of containment airborne activity in the drywell and suppression chamber air volume to drywell mixing only for the first 2 hours and then mixing in the drywell and suppression chamber air volumes instantaneously at 2 hours (Table 1E, Item 1-1),

6) not crediting a reduction in containment atmosphere pressure for a reduction in containment and MSIV leakage after 24 hours (Table IE, Items II-1 and V-I), and

7) reduction in the control room unfiltered inleakage from 350 cfm to 300 cfm (Table IE, Item II-1).

Revisions 2 and 4 of this calculation used the J. E. Cline model for calculating natural removal of elemental iodine from the main steam line compartment volumes. Revision 3 of this calculation applies a decontamination factor of 2 to account for natural removal of elemental iodine from each main steam line compartment volume. The factor of 2 decontamination factor is from Appendix B of AEB-98-03.

CALCULATION NO. H-1-ZZ-MDC-1880 I REVISION NO. 4 PAGE NO.53 of 117 Application of the J. E. Cline model results in larger elemental iodine releases for the first 6 hours post LOCA than would be attained by applying a decontamination factor of 2 (Refer to Tables IL and IM). Application of a decontamination factor of 2 for post LOCA times greater than 6 hours would result in larger elemental iodine releases from the main steam line compartment volumes than would be attained using the J. E. Cline model. Credit for elemental iodine removal from main steam line pipe compartment volumes is terminated for post LOCA times greater than 96 hours in Revisions 3 and 4. Use of the J. E. Cline model for calculating EAB doses due to MSIV leakage, as opposed to using the decontamination factor of 2 approach, will result in larger EAB doses since the EAB dose is limited to the first few hours post LOCA. Application of the J. E. Cline model is therefore conservative with respect to calculation of EAB doses due to MSIV leakage. Use of the J. E. Cline model for calculating the 30 day post LOCA control room and LPZ doses due to MSIV leakage, as opposed to using the decontamination factor of 2 approach, is expected to be slightly nonconservative with respect to the dose contribution attributed to MSIV leakage, but the differences in doses obtained using the 2 different models are expected to be small. The following is a comparison of the dose contributions in Revisions 2, 3, and 4 to the control room, the EAB, and the LPZ doses from the three release pathways and an assessment of the impact that changes in input parameters and changes in methodology have on control room, EAB, and LPZ doses for the three release pathways. Comparison of Revision 2, 3, and 4 Doses _ _ Post LOCA TEDE Dose (Rem) Containment ESF Leakage MSIV Leakage Total Dose ( Leakage Revision 2 1.01E+00 1.17E+00 1.97E+00 4.16E+00 Control Room Revision 3 4.79E-01 3.33E+00 3.31E-01 4.17E+00 Revision 4 5.37E-01 2.88E+00 6.13E-01 4.04E+00 Revision 2 4.25E-01 1.80E-01 2.49E+00 3.10E+00 EAB Revision 3 3.87E-01 5.22E-01 5.01E-01 1.43E+00 Revision 4 3.87E-01 5.22E-01 8.68E-01 1.78E+00 Revision 2 1.69E-01 9.19E-02 4.35E-01 6.96E-01 LPZ Revision 3 1.31E-01 2.64E-01 1.51E-01 5.48E-01 Revision 4 1.47E-01 2.64E-01 2.63E-01 6.74E-01 (1) In addition to the dose contributions from the three release pathways (i.e., containment leakage, ESF leakage, and MSIV leakage) the control room dose includes a dose contribution due to shine from the control room filter. The contribution from the control room filter to the control room dose is 1.03E-02 rem, 1.26E-02 rem, and 1.29E-02 rem for Revisions 2, 3, and 4, respectively. ESF Leakage Release Pathway Revision 2 to Revision 3 The only input parameter that changed between Revision 2 and Revision 3 and that impacts the doses due to releases from this release pathway is the ESF leak rate that was changed from 1 gpm to 2.85 gpm. Doses at the EAB, the LPZ, and in the control room are expected to increase in direct proportion to the increase in ESF leak rate (i.e., to increase by the ratio of 2.85/1). The doses calculated in Revision 2 from the ESF leakage pathway are 1.17, 0.180, and 0.0919 rem for the control room, the EAB, and the LPZ, respectively. The corresponding doses in Revision 3 are 3.33, 0.522, and 0.264 rem. A

CALCULATION NO. H-1-ZZ-MDC-1880 I REVISION NO. 4 IPAGE NO. 54 of 117 comparison of the dose results from Revision 3 to dose results of Revision 2 indicates that the doses do increase in direct proportion to the ESF leak rate. Revision 3 to Revision 4 The input parameter that changed between Revision 3 and Revision 4 which impacts the doses due to releases from this release pathway is the control room unfiltered inleakage rate which was reduced from 350 cfm to 300 cfm. The ESF leakage rate remains at 2.85 gpm. A reduction in the control room unfiltered inleakage rate results in a smaller dose to the control room, but does not affect the EAB or LPZ doses due to releases via the ESF leakage pathway. Section 7.6.1 of this calculation shows that the unfiltered inleakage accounts for over 95% of the radioactivity that gets into the control room. As such, the control room dose is nearly directly proportional to the unfiltered inleakage rate and the control room dose impact associated with reducing the control room unfiltered inleakage rate from 350 cfm to 300 cfm is expected to result in a control room dose reduction of approximately 14% (i.e., 50 cfm/350 cfm). The doses calculated in Revision 3 from the ESF leakage pathway are 3.33, 0.522, and 0.264 rem for the control room, the EAB, and the LPZ, respectively. The corresponding doses in Revision 4 are 2.88, 0.522, and 0.264 rem. A comparison of the dose results from Revision 4 to dose results of Revision 3 indicates that the EAB and LPZ doses remain unchanged from the doses calculated in Revision 3 and the control room dose is reduced by 0.45 rem (i.e., 3.33 - 2.88) which is approximately 14% (i.e., 100 x 0.45/3.33). Revision 2 to Revision 4 The input parameters that changed between Revision 2 and Revision 4 which impact the doses due to releases from this release pathway are the ESF leak rate which changed from I gpm in Revision 2 to 2.85 gpm in Revision 4 and the control room unfiltered inleakage rate which was reduced from 350 cfm in Revision 2 to 300 cfm in Revision 4. EAB and LPZ doses are expected to increase in direct proportion to the increase in ESF leak rate. The reduction in the control room unfiltered inleakage rate reduces the effect of the increase in ESF leak rate on the control room dose. As previously indicated, unfiltered inleakage accounts for the majority of the dose to the control room. The combined impact of an increase in the ESF leak rate and a reduction in the control room unfiltered inleakage rate on the control room doses is proportional to the product of the ratio of the ESF leak rates (i.e., 2.85/1) and the ratio of the control room unfiltered rate (i.e., 300/350). The doses calculated in Revision 2 from the ESF leakage pathway are 1.17, 0.18, and 0.0919 rem for the control room, the EAB, and the LPZ, respectively. The corresponding doses in Revision 4 are 2.88, 0.522, and 0.264 rem. A comparison of the dose results from Revision 4 to dose results of Revision 3 indicates that the EAB and LPZ doses increase in direct proportion to the ESF leak rate and the control room dose increases in proportion to the product of (2.85/1) x (300/350). Containment Leakage Release Pathway Revision 2 to Revision 3 The input parameters that changed between Revision 2 and Revision 3 which impact the doses due to releases from this release pathway are changes in the containment air dilution volume and credit for plateout of elemental iodine on wetted containment surfaces. In Revision 2, activity released into containment is immediately mixed within the drywell and suppression chamber air volumes (306000 ft3). In Revision 3, activity released into containment during the first 2 hours of the LOCA is mixed within the drywell air volume (169000 ft3). After 2 hours airborne activity is mixed within the drywell and suppression chamber air volumes. Limiting the dilution volume for containment airborne sources to the drywell air volume for the first 2 hours, increases the concentration of airborne radioactivity within the drywell air which increases the amount of activity released into the environment. Therefore, changes in the containment air dilution volume result in increased dose contributions for the first 2 hours

CALCULATION NO. H-1-ZZ-MDC-1880 I REVISION NO. 4 1 PAGE NO. 55 of 117 following the LOCA. Credit for plateout of elemental iodine on wetted containment surfaces reduces the amount of airborne activity in containment atmosphere that is available for release which reduces the amount of activity released to the environment. Elemental iodine is removed from containment atmosphere onto containment wetted surfaces until a decontamination factor of 200 is reached which occurs at approximately 4 hours into the LOCA. The reduction in EAB, LPZ and control room doses associated with the reduction in containment airborne activity due to elemental iodine removal via plateout on containment wetted surfaces is larger than the increase in doses associated with increased radioactivity releases during the first 2 hours due to limiting airborne activity to the drywell volume. The overall effect is a reduction in control room, EAB, and LPZ doses. The doses calculated in Revision 2 from the containment leakage pathway are 1.01, 0.425, and 0.169 rem for the control room, the EAB, and the LPZ, respectively. The corresponding doses in Revision 3 are 0.479, 0.387, and 0.131 rem. A comparison of the dose results from Revision 3 to dose results of Revision 2 indicates that the control room, EAB, and LPZ doses all decrease. Revision 3 to Revision 4 The input parameters that changed between Revision 3 and Revision 4 which impact the doses due to releases from this release pathway are the control room unfiltered inleakage rate which was reduced from 350 cfm to 300 cfm, and in Revision 4 the containment leak rate was held constant during the course of the accident whereas in Revision 3, the containment leak rate was reduced by a factor of 2 at 24 hours. Since the containment leak rate is not reduced by a factor of 2 at 24 hours more activity is leaked from the containment building after 24 hours and so the dose contributions from containment leakage after 24 hours would increase for Revision 4. Since the EAB dose is acquired within the first few hours following the LOCA, there is not expected to be a significant change to the EAB dose in Revisions 4 from the dose in Revision 3. This is observed by comparing the dose results between Revision 3 and Revision 4. Specifically, the EAB dose (0.387 Rem) which is acquired within 24 hours is the same for Revisions 3 and 4. The control room and LPZ doses calculated in Revision 3 from the containment leakage pathway are 0.479 and 0.131 rem, respectively. The corresponding doses in Revision 4 are 0.537 and 0.147 rem. The control room and LPZ doses in Revision 4 are larger than the doses in Revision 3 by approximately 12%. This increase is attributed to the additional release of activity from containment atmosphere after 24 hours since a containment leak rate reduction factor of 2 was not applied to Revision 4. The increase in the control room dose is slightly smaller than the increase in the LPZ dose which is attributed to a reduction in the control room unfiltered inleakage. Revision 2 to Revision 4 The input parameters that changed between Revision 2 and Revision 4 which impact the doses due to releases from this release pathway are changes in the containment air dilution volume, credit for plateout of elemental iodine on wetted containment surfaces, changes in the control room unfiltered inleakage rate which was reduced from 350 cfm to 300 cfm and changes in the containment leak rate. In Revision 4 the containment leak rate was held constant during the course of the accident whereas in Revision 2 the containment leak rate was reduced by a factor of 2 at 24 hours. In Revision 2, activity released into containment is immediately mixed within the drywell and suppression chamber air volumes whereas in Revision 4 radioactivity released in the containment is confined to the drywell air volume for the first 2 hours and after 2 hours airborne activity is mixed within the drywell and suppression chamber air volumes. Changes in the containment air dilution volume and the containment leak rate increase radioactivity releases via the containment leakage pathway in Revision 4 while crediting for, elemental iodine plateout on wetted containment surfaces reduce radioactivity releases via this leakage pathway. The reduction in the control room unfiltered inleakage rate in Revision 4 reduces the amount of radioactivity that enters the control room. The doses calculated in Revision 2 from the containment leakage pathway are 1.01, 0.425, and 0.169 rem for the control room, the EAB, and the LPZ, respectively. The corresponding doses in Revision 4 are 0.537, 0.387, and 0.147 rem. The effect of the

CALCULATION NO. H-1-ZZ-MDC-1880 I REVISION NO. 4 I PAGE NO. 56 of 117 changes made in Revision 4 on the doses at the EAB, LPZ and control room are dominated by the reduction in radioactivity releases due to plateout of elemental iodine on containment wetted surfaces. The combined effect of the changes is a reduction in the EAB, the LPZ, and the control room doses. The control room dose is reduced the most in Revision 4 because the reduction in unfiltered inleakage reduces the amount of radioactivity that enters the control room compared to Revision 2. MSIV Leakage Release Pathway Revision 2 to Revision 3 Changes between Revision 2 and Revision 3 which are expected to have significant impact on doses due to releases from this pathway include changes in the MSIV leakage volumetric flow rates, changes in the number of main steam line compartment volumes considered in both the failed and intact main steam lines, use of plug flow model with credit for holdup in Revision 2 versus a well mixed flow model with no credit for holdup in Revision 3, differences in determination of pipe surface area for aerosol deposition, and taking credit for elemental iodine plateout on containment wetted surfaces. The MSIV leakage volumetric flow rate in Revision 2 is the measured flow rate at standard temperature and pressure conditions whereas the MSIV leakage volumetric flow in Revision 3 is the standard measured volumetric flow rate adjusted for containment post LOCA temperature and pressure conditions. It is indicated in Section 7.2 of this calculation that the volumetric flow rate at post LOCA temperature and pressure conditions is approximately a third of the measured volumetric flow rate at standard temperature and pressure conditions for the first 2 hours post LOCA and a fifth of measured volumetric flow rates thereafter. The amount of radioactivity released through the MSIV pathway is directly proportional to the volumetric flow rate and therefore the change in determination of volumetric flow rate in Revision 3 is expected to result in the greatest change in doses and result in dose reductions by approximately a factor of 3. The failed and intact main steam lines in Revision 2 are each treated as a single pipe volume for determining aerosol deposition and elemental iodine removal. In Revision 3 the failed and intact main steam lines are both modeled as two separate compartment volumes in series for calculating aerosol deposition and elemental iodine removal. The use of two pipe volumes in series compounds the removal efficiency for aerosols and elemental iodine. Use of two pipe volumes in series as opposed to one, is expected to result in the second greatest reduction in Revision 3 doses via the MSIV release pathway. All of the identified changes between Revision 2 and Revision 3 cause a reduction in radioactivity releases via the MSIV leakage release pathway and thus a reduction in doses. Changes in volumetric flow rate and number of pipe volumes dominate dose impacts and the other changes made between Revision 2 and Revision 3 have secondary effects on doses. The doses calculated in Revision 2 from the MSIV leakage pathway are 1.97, 2.49, and 0.435 rem for the control room, the EAB, and the LPZ, respectively. The corresponding doses in Revision 3 are 0.331, 0.501, and 0.151 rem. A comparison of the dose results from Revision 2 to Revision 3 indicates that the control room dose was reduced by approximately a factor of 6, the EAB dose was reduced by approximately a factor of 5 and the LPZ dose was reduced by approximately a factor of 3. Revision 3 to Revision 4 Changes between Revision 3 and Revision 4 which impact the doses due to releases from this release pathway are changes in the control room unfiltered inleakage rate which was reduced from 350 cfm to 300 cfm, changes in the containment leak rate, and changes in the number of main steam line compartment volumes considered in both the failed and intact main steam lines. The failed and intact main steam lines were each modeled as a single compartment volume in Revision 4 whereas they were modeled as two compartment volumes in series for each the failed and intact main steam line for Revision 3. In Revision 4 the MSIV leak rate was held constant during the course of the accident whereas in Revision 3 the MSIV leak rate was reduced by a factor of 2 at 24 hours. The reduction in control room unfiltered inleakage in Revision 4 reduces the amount of activity that enters the control

CALCULATION NO. H-1-ZZ-MDC-1880 I REVISION NO. 4 1 PAGE NO. 57 of 117 room and thus it reduces the control room dose. The unfiltered inleakage accounts for over 95% of the radioactivity that gets into the control room. As such, the control room dose is nearly directly proportional to the unfiltered inleakage rate and the control room dose impact associated with reducing the control room unfiltered inleakage rate from 350 cfm to 300 cfm is expected to result in a control room dose reduction of approximately 14% (i.e., 50 cfm/350 cfm). The remaining changes that were made in Revision 4 result in greater radioactivity releases via the MSIV leakage pathway and therefore larger doses to the control room, to the EAB, and to the LPZ, than the doses of Revision 3. The doses calculated in Revision 3 for the MSIV leakage pathway are 0.331, 0.501, and 0.151 rem for the control room, the EAB, and the LPZ, respectively. The corresponding doses in Revision 4 are 0.613, 0.868, and 0.263 rem. A comparison of the dose results from Revision 3 to Revision 4 indicate that the overall increase in the control room dose is approximately 80% (from 0.331 Rem to 0.613 Rem). The EAB and LPZ doses increased by approximately 73% and 74%, respectively. Revision 2 to Revision 4 Changes between Revision 2 and Revision 4 which are expected to result in significant reductions in doses due to releases from this pathway include changes in the MSIV leakage volumetric flow rates, credit for removal of elemental iodine from containment atmosphere onto wetted containment surfaces, and changes in the control room unfiltered inleakage rate which was reduced from 350 cfm to 300 cfm. The MSIV leakage volumetric flow rate in Revision 2 is the measured flow rate at standard temperature and pressure conditions whereas the MSIV leakage volumetric flow in Revision 3 is the standard measured volumetric flow rate adjusted for containment post LOCA temperature and pressure conditions. As indicated in the discussion for differences between Revision 2 and 3 for this release pathway, the change in determination of volumetric flow rate in Revision 4 is expected to result in dose reductions by approximately a factor of 3. The reduction in control room unfiltered inleakage in Revision 4 reduces the amount of activity that enters the control room and thus it reduces the control room dose by approximately 14% as indicated in the discussion of differences between Revision 3 and Revision 4 for this release pathway. Changes between Revision 2 and Revision 4 which are expected to result in significant increases in control room, EAB, and LPZ doses due to releases from this pathway include changes in MSIV leak rate. In Revision 4 the MSIV leak rate is held constant during the course of the accident whereas in Revision 2 the MSIV leak rate was reduced by a factor of 2 at 24 hours. The change in the MSIV leak rate results in larger doses in Revision 4 for the LPZ (which receives dose contributions during the entire accident) but it will not affect the EAB dose which is limited to a few hours after the accident. The doses calculated in Revision 2 from the MSIV leakage pathway are 1.97, 2.49, and 0.435 rem for the control room, the EAB, and the LPZ, respectively. The corresponding doses in Revision 4 are 0.613, 0.868, and 0.263 rem. A comparison of the dose results from Revision 2 to Revision 4 indicates that the control room dose was reduced by slightly more than a factor of 3, the EAB dose was reduced by slightly less than a factor of 3, and the LPZ dose was reduced by less than a factor of 2. The dose reduction in Revision 4 is dominated by the change in determination of volumetric flow rate. The larger reduction in the control room dose compared to the LPZ dose is attributed primarily to reduced unfiltered inleakage into the control room. Impact of Changes on Combined Doses from All Release Pathways Revision 2 to Revision 3 A comparison of the Revision 3 control room, EAB, and LPZ doses for the ESF leakage pathway to the Revision 2 doses indicates that the doses increase significantly in Revision 3. The increase in doses is approximately a factor of 2.85 and is directly proportional to the increase in ESF leakage which is increased from I gpm to 2.85 gpm. The increase in control room doses due to ESF leakage (i.e., an increase of 2.16 rem (3.33 - 1.17) to the control room) is offset by the reduction in doses from the containment leakage pathway (i.e., 0.531 rem (1.01 - 0.479)) and MSIV leakage pathway (i.e., 1.639

CALCULATION NO. H-1-ZZ-MDC-1880 I REVISION NO. 4 1 PAGE NO. 58 of 117 rem (1.97 - 0.33 1)). The reduction in the control room, EAB, and LPZ doses in the containment leakage pathway is primarily attributed to crediting for reduction in airborne elemental iodine from the containment atmosphere due to plateout on wetted containment surfaces. The reduction in the control room, EAB, and LPZ doses due to releases via the MSIV leakage pathway is primarily attributed to using post LOCA drywell temperature and pressure to determine MSIV volumetric flow rate instead of using the volumetric flow rate associated with standard temperature and pressure conditions. Determining MSIV releases using volumetric flow rates based on post LOCA drywell temperature and pressure conditions reduces the doses by approximately a factor of 3. Other contributors to the reduction in doses via the MSIV leakage pathway are the reduction in airborne elemental iodine from the containment atmosphere due to plateout on wetted containment surfaces and the use of two compartment volumes in series on both the failed and intact main steam lines for Revision 3. In Revision 2, only one compartment volume was used for each of the failed and intact main steam lines. Revision 2 to Revision 4 A comparison of the Revision 4 control room, EAB, and LPZ doses for the ESF leakage pathway to the Revision 2 doses indicates that the doses increase significantly in Revision 4. The increase in EAB and LPZ doses is approximately a factor of 2.85 and is directly proportional to the increase in ESF leakage which is increased from 1 gpm to 2.85 gpm. The increase in the control room dose is less than a factor of 2.85 (i.e., an increase by a factor of 2.46 (2.88/1.17)) and is attributed to a reduction in the control room unfiltered inleakage rate from 350 cfm to 300 cfm in Revision 4. The increase in doses due to ESF leakage (i.e., an increase of 1.71 rem (2.88 - 1.17) to the control room) is offset by the reduction in doses from the containment leakage pathway (i.e., 0.473 rem (1.01 - 0.537)) and MSIV leakage pathway (i.e., 1.357 rem (1.97 - 0.613)). The reduction in the control room, EAB, and LPZ doses in the containment leakage pathway is primarily attributed to crediting for reduction in airborne elemental iodine from the containment atmosphere due to plateout on wetted containment surfaces. The reduction in Revision 4 containment leakage pathway doses compared to Revision 2 doses is not as great as the reduction attained in Revision 3 containment leakage pathway doses compared to Revision 2 doses because Revision 4 doesn't credit a 50% reduction in containment leakage (at 24 hours or any other time post LOCA) due to reduced containment pressure. The reduction in the control room, EAB, and LPZ doses due to releases via the MSIV leakage pathway is primarily attributed to using post LOCA drywell temperature and pressure to determine MSIV volumetric flow rate instead of using the volumetric flow rate associated with standard temperature and pressure conditions. Determining MSIV releases using volumetric flow rates based on drywell post LOCA temperature and pressure conditions reduces the doses by approximately a factor of 3. Another contributor to the reduction in doses via the MSIV leakage pathway is the reduction in airborne elemental iodine from the containment atmosphere due to plateout on wetted containment surfaces. Note, Revision 4 modeled a single compartment volume for both the failed and intact main steam lines. Revision 2 also used single compartment volumes and so there was no dose reduction due to use of two compartment volumes in series as there was for Revision

3. The compounding effect that two main steam line volume compartments in series have oh aerosol deposition and elemental iodine removal account for the much smaller control room, EAB, and LPZ doses seen in Revision 3 for the MSIV pathway releases.

9.0 CONCLUSION

S The Section 8.1 results, indicate that the EAB, LPZ, and CR doses are within their allowable TEDE limits for the increased ESF leakage rate up to 2.85 gpm and the updated MSIV leakage model. The comparisons in Section 8.2 show that the CR, EAB and LPZ total calculated doses are less than previously calculated values.

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO. 4 PAGE NO.59 of 117

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-I -ZZ-MDC-1 879, 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.I.c.1, FRVS Vent HEPA Filter Testing Criterion 10.6.2 Specification 3/4.6.6 Primary Containment Atmosphere Control 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 3/4.6.2, 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 10.6.13 Specification 4.7.2.1 .c. 1, Control Room Emergency Filtration System Surveillance Requirements 10.6.14 Specification 4.7.2.1 .c.2, Control Room Emergency Filtration System Surveillance Requirements 10.6.15 Specification 4.7.2.1 .c.3, Control Room Emergency Filtration System Surveillance Requirements 10.6.16 Specification 4.7.2.1.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 Systemn 1 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.

CALCULATION NO. H-I-ZZ-MDC-1880 I REVISION NO. 4 TPAGE NO. 60 of 117 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: 10.12.a 1-P-AB-001, Rev 11 10.12.b 1-P-AB-002, Rev 9 10.12.c 1-P-AB-003, Rev 9 10.12.d 1-P-AB-004, Rev 9 10.12.e 1-P-AB-011, Rev 11 10.12.f 1-P-AB-01, Rev 18 10.13 Piping Area Drawings: 10.13.a P-1703-1, Rev 3, Reactor Building Area 17, Plan EL 100'-2" 10.13.b P-1704-1, Rev 2, Reactor Building Area 17, Plan EL 112'-0" 10.13.c P-1705-1, Rev 2, Reactor Building Area 17, Plan EL 121'-7-1/2" 10.13.d P-1712-1, Rev 2, Reactor Building Area 17, Section B17 - B17 10.13.e P-1713-1, Rev 4, Reactor Building Area 17, Section C17 - C17 10.13.f P-1403-1, Rev 2, Reactor Building Area 14, Plan At EL 102'-0" 10.13.g P-1414-l, Rev 1, Reactor Building Area 14, Section D14 - D14 10.13.h P-1304-1, Rev 4, Piping Area Drawing, Reactor Building Area 13, Plan at El. 132'-0"

                      & El. 145' 10.13.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 10.1 4.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 Vendor Technical Document (VTD) No. 430024, Volume 002, EPU TR T0400 - Containment System Response. 10.18 NUREG-0800, Standard Review Plan, "Radiological Consequence Analyses Using Alternative Source Terms," SRP 15.0.1, Rev. 0, July 2000. 10.19 Calculation No. GU-0013, Rev. 4, Filtration Recirculation and Ventilation System Exhaust Rate 10.20 GE-Hitachi Report NEDC-33529P, "Safety Analysis Report to Support Introduction of GE 14i Isotope Test Assemblies (ITAs) in Hope Creek Generating Station," Revision 0, December 2009.

CALCULATION NO. H-1-ZZ-MDC-1880 I REVISION NO. 4 PAGE NO. 61 of 117 10.21 Main Steam Isometric Drawings: 10.21.a. FSK-P-214, Rev 14, Main Steam A & B Inside Drywell 10.21.b. FSK-P-215, Rev 15, Main Steam C & D Inside Drywell 10.22 NRC Report AEB-98-03, "Assessment of Radiological Consequences For the Perry Pilot Plant Application Using the Revised (NUREG-1465) Source Term. 10.23 U.S. NRC Regulatory Guide 1.197, Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors, May 2003. 10.24 Hope Creek Calculation No. 12-102(Q), Rev 0, Surface Area Inside Drywell 10.25 Calculation No. No. H-1-ZZ-MDC-0364, Rev 1, 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: 10.27.a P-0006-0, Rev 7, Plan EL 153'-0" & 162'-0" 10.27.b P-0011-0, 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: 10.31 .a C-1413-0, Rev 20, Floor Plan EL 146'-0", EL 150'-0", EL 155'-3" Area 27 10.31.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 PSEG Procedure No. LS-AA-104-1000, Revision 3, 50.59 Resource Manual.

CALCULATION NO. H-l-ZZ-MDC-1880 I REVISION NO. 4 1 PAGE NO. 62 of 117 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 NUREG-0800, Standard Review Plan, "Containment Spray as a Fission Product Cleanup System," SRP 6.5.2, Revision 2, 1988. 10.42 Hope Creek License Amendment 160, RE: Elimination of Requirements for Hydrogen Recombiners and Hydrogen/Oxygen Monitors Using Consolidated Line Item Improvement Process (TAC No. MC4792) 10.43 Calculation No. H-1-ZZ-MDC-1886, Rev 0, Hope Creek Post-Accident pH. 10.44 Nuclear Energy Institute Report No. NEI 96-07, Rev 1, Guidelines for 10 CFR 50.59 Implementation. 10.45 Vendor Technical Document (VTD) No. 430058, Volume 002, Rev 1, EPU TR T0802, Radioactive Source Term - Core Inventory. 10.46 NCS Corporation, "Control Room Envelope Inleakage Testing At Hope Creek Generating Station", December 28, 2009. 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 2, RADTRAD Computer Code. 10.49 NRC Safety Evaluation Report, Hope Creek Generating Station - Issuance of Amendment No. 134 for Increase in Allowable MSIV Leakage Rate and Elimination of MSIV Sealing System. 10.50 NRC letter to PSEG Nuclear dated April 15, 2003, "Hope Creek Generating Station - Issuance of Amendment 146 Re: Containment Requirements During Fuel Handling and Removal of Charcoal Filters (TAC No. MB5548)." 10.51 Not Used. 10.52 MSIV Leakage Iodine Transport Analysis By J.E. Cline & Associates, March 26, 1991, Contract NRC-03-87-029, Task Order 75 10.53 HCGS Containment Atmosphere Control P & ID M-57-1, Revision 41 (Sheet 1 of 2). 10.54 HCGS Technical Requirements Manual, Revision 2 10.55 "NRC Regulatory Issue Summary 2006-04 Experience with Implementation of Alternative Source Terms," RIS 2006-04, US NRC, March 7, 2006 (ADAMS Accession No. ML053460347). 10.56 NUREG-0800, Standard Review Plan, Branch Technical Position 6-4 "Containment Purging During Normal Plant Operations," BTP 6-4, Revision 3, March 2007.

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO. 4 1 PAGE NO. 63 of 117 10.57 "NRC Regulatory Issue Summary 2001-19 Deficiencies in the Documentation of Design Basis Radiological Analyses Submitted in Conjunction with License Amendment Requests," RIS 2001-19, US NRC, October 18, 2001 (ADAMS Accession No. ML011860407).

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO. 4 PAGE NO.64 of 117 11.0 TABLES Table IA Discharge Bundle Inventory Including Parent/Daughter Isotopes Per Reference 10.45, Appendix C 1?nd af ('vdl fDigphnrop Ruindle Inventnrv 2t ,*hiutdnwn Discharge Bundle Discharge Bundle Discharge Bundle Discharge Bundle Isotope Inventory Inventory Isotope Inventory Inventory (Ci/MWt) (Ci/MWt) (Ci/MWt) (Ci/MWt) CO-58* 1.529E+02 1.529E+02 TE43 I& 1&i 2.4611- 04<;' CO-60* 1.830E+02 1.830E+02 ITE 13I MN 4.15 3E--li3 Iý 2.876E+o4 KR-85 4.711EE+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 1 1 71 ( SR-90 3.790E+03 3.790E+03 N i5 iE04 1.I

             ,,R-916771#Ou                      4 .23 1E+/-0 I           XE-133           5.306E+04            5.306E+04
      ... 91j4...         I 554f 04                                    XE-135            1.482E+04            1.482E+04 SR-92              2.990E+04              2.990E+04             CS-134            1.319E+04            1.319E+04 Y-90             3.981E+03              3.981E+03             CS-136           3.704E+03             3.704E+03 Y-91             2.750E+04              2.750E+04             C    _137* *\.6,261+                   I 0'96E.O4.

Y-92 3.005E+04 3.005E+04 BAI 37M 5.3':K*A9E+3E)t,3 Y-93 3.607E+04 3.607E+04 BA-139 4.760E+04 4.760E+04 ZR-95 4.217E+04 4.217E+04 BA-140 4.590E+04 4.590E+04 ZR-Q97 <4.4191. o4 1.3E071 oLA- 140 4.981E+04 4.981E+04 N13-97M\ 4.1911E+/-{o LA-141 4.325E+04 4.325E+04 NB,1-9-7 4,164EA:04 'LA- 142 4.134E+04 4.134E+/-04 NB-95 4.237E+04 4.237E+04 CE-141 4.350E+04 4.350E+04 MO-99 5.278E+04 5.278E+04 CE- 143 3.910E+04 3.910E+04 TC-99M 4.62]E+04 4.621E+04 PU-_,39 3.3338E+/- 41 -. 34E3031 RU-105 3.529E+04 3.529E+04 PR-143 3.783E+04 3.783E+04 RU-0o ý9-Oz 4722+0 ND-147 1.783E+04 1.783E+04 i RFJI-1l06 463E+0 NP-239 6.917E+05 6.917E+05 RB- 105 3.237E+04 3.237E+04 PU-238 3.442E+02 3.442E+02 SB-127 3 .379E+03 3.379E+03 PU-239 1.333E+01 1.333*E01 SB-129 9.569E+03 9.569E+03 PU-240 2.675E+01 2.675E+01 TE-127 3.355E+03 3.355E+03 PU-241 5.419E+03 5.419E+03 TE-127M 4.508E+02 4.508E+02 AM-241 7.266E+00 7.266E+00 TE-129 9.430E+03 9.430E+03 CM-242 2.567E+03 2.567E+03 TE-129M 1.401E+03 1.401E+03 CM-244 5.188E+02 5.188E+02

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

ICALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO.4 PAGE NO.65 of 117 Table 1B Core Inventory - Discharge Bundle Inventory Discharge Discharge Bundle Bundle Isotope Inventory Isotope Inventory (Ci/MWt) (CitMWt CO-58* 1.529E+02 TE-131M 2.876E+04 CO-60* 1.830E+02 TE-132 3.917E+04 KR-85 4.711E+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-133 5.306E+04 SR-90 3.790E+03 XE-135 1.482E+04 SR-91 4.231EE+04 CS-134 1.319E+04 SR-92 2.990E+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 4.981E+04 ZR-95 4.217E+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 4.621 E+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 1.401E+03 CM-242 2.576E+03 TE- 129 9.430E+03 CM-244 5.188E+02 Total Core Inventory From Table 1A

CALCULATION NO. H-1-ZZ-MDC-1880 7 REVISION NO. 4 PAGE NO. 66 of 117 Table IC Higher Core Inventory Including Parent/Daughter Isotopes Per Reference 10.46, Appendix A End of Cycle Core Average Inventory at Shutdown Core Core Core Core Inventory Inventory Inventory Inventory Isotope Isotope (Ci/MWt) (CiIMWt) (Ci/MWt) (CiIMWt) CO-58* 1.529E+02 1.529E+02 ...1 1*.390E÷+ *4 CO-60* 1.830E+02 1.830E+02 I'- N1 ',3840E-03; 2.764E+04 KR-85 3.330E+02 3.300E+02 TE-132 3.810E+04 3.810E+04 KR-85M 7.350E+03 7.350E+03 1-131 2.670E+04 2.670E+04 KR-87 1.420E+04 1.420E+04 1-132 3.870E+04 3.870E+04 KR-88 2.OOOE+04 2.OOOE+04 1-133 5.510E+04 5.510E+04 RB-86 6.350E+01 6.350E+01 1-134 6.060E+04 6.060E+04 SR-89 2.690E+04 2.690E+04 MIsT < 5 50EOA(*2,

• 60E4-04 SR-90 2.640E+03 2.640E+03 IK20* 0103

         *S'R~

• 6 0 ~ LAXE-133 5.300E+04 5.300E+04

               \It7 Q     I1ý45004                                 XE-135               1.820E+04       1.820E+04 SR-92           3.610E+04              3.610E+04          CS-134              5.350E+03      5.350E+03 Y-90            2.8112.810OE+03                           CS-136              1.860E+03       1.860E+03 Y-91            3.440E+04              3.440E+04         N50E+05                             o.2203 Y-92            3.620E+04              3.620E+04         PU-22890E+O+

Y-93 4.160E+04 4.160E+04 BA-139 4.950E+04 4.950E+04 ZR-95 4.850E+04 4.850E+04 BA-140 4.780E+04 4.780E+04 TBE-7M 71 3.1E LA- 141 4.510E+04 4.510E+04

         ',ýB-9           ý.5* E004*:;                             LA- 142              4.3 70E+04     4.370E+04 NB-95            4.870E+04              4.870E+04         CE- 141              4.540E+04      4.540E+04 M 0-99            5. 100E+04             5.100OE+04        CE- 143              4.220E+04      4.220E+04 TC-99M            4.460E+04              4.460E+04               144*,i      *368E           g{ .;4Z4*

RU-105 2.700E+04 2.700E+04 PR-143 4.080E+04 4.080E+04 TE-RI 29M0 1.210F+0 IIND14 1.810E+04 1.810E+04 R1 I- 1005,1 E3 4 IN NP-239 5.220E+05 5.220E+05 RH-105 2.530E+04 2.530E+04 PU-238 9.040E+01I .040E+01I SB-127 2.800E+03 2.800E+03 'PU-239 1.090E+01 1.090E+01 SB- 129 8.490E+03 8.490E+03 PU-240 1.410OE+0 1 1.410OE+01I TE-127 2.780E+03 2.780E+03 PU-241 4.090E+03 4.090E+03 TE- 127M 3.7 10E+02 3.710OE+02 AM-241 4.600E+00 4.600E+00 TE-129 8.350E+03 8.350E+03 CM-242 1.090E+03 1.090E+03 TE- I129M 1.240E+03 1.240E+03 CM-244 5.240E+01I 5.240E+01

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

CALCULATION NO. H-1-ZZ-MDC-1880 I REVISION NO. 4 1PAGE NO.67 of 117 Table 1D Core Inventory - Core Average Exposure Total Total Core Core Isotope Inventory Isotope Inventory (Ci/MWt) (Ci/MWt) CO-58* 1.529E+02 TE-131M 2.764E+04 CO-60* 1.830E+02 TE-132 3.810E+04 KR-85 3.330E+02 1-131 2.670E+04 KR-85M 7.350E+03 1-132 3.870E+04 KR-87 1.420E+04 1-133 5.510E+04 KR-88 2.OOOE+04 1-134 6.060E+04 RB-86 6.350E+01 1-135 6.220E+04 SR-89 2.690E+04 XE-133 5.300E+04 SR-90 2.640E+03 XE-135 1.820E+04 SR-91 5.300E+04 CS-134 5.350E+03 SR-92 3.610E+04 CS-136 1.860E+03 Y-90 2.810E+03 CS-137 6.760E+03 Y-91 3.440E+04 BA-139 4.950E+04 Y-92 3.620E+04 BA-140 4.780E+04 Y-93 4.160E+04 LA-140 5.080E+04 ZR-95 4.850E+04 LA-141 4.510E+04 ZR-97 1.468E+05 LA-142 4.370E+04 NB-95 4.870E+04 CE-141 4.540E+04 MO-99 5.100E+04 CE- 143 4.220E+04 TC-99M 4.460E+04 CE- 144 7.424E+04 RU- 103 7.700E+04 PR-143 4.080E+04 RU-105 2.700E+04 ND-147 1.810E+04 RU-106 2.940E+04 NP-239 5.220E+05 RH-105 2.530E+04 PU-238 9.040E+01 SB-127 2.800E+03 PU-239 1.090E+01 SB-129 8.490E+03 PU-240 1.410E+0O1 TE-127 2.780E+03 PU-241 4.090E+03 TE-127M 3.7101E+02 AM-241 4.600E+00 TE-129 8.350E+03 CM-242 1.090E+03 TE- 129M 1.240E+03 CM-244 5.240E+O1 Total Core Inventory From Table IC

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO. 4 PAGE NO. 68 of 117 Table IE Comparison of Modeling Differences between Revision 2, Revision 3, and Revision 4 of Calculation H-1-ZZ-MDC-1880 Item Changed Parameter Calc H-I-ZZ-MDC-1880, Calc H-1-ZZ-MDC- Calc H-1-ZZ-MDC-1880, Type of Change

      #                                            Revision 2               1880, Revision 3              Revision 4         Input or Methodology

1. Airborne Radioactivity in Containment Atmosphere I-1 Airborne Activity Dilution Radioactivity released from Radioactivity released Same as Revision 3. Methodology in Containment the fuel mixes from core is diluted in (See Section 2. 1.1) instantaneously and drywell air volume for homogeneously in first 2 hours of the LOCA containment free air volume and then in combined which consists of drywell drywell plus suppression and suppression chamber chamber air volumes.

air volumes. 1-2 Elemental Iodine Removal No credit taken for Removal of elemental Same as Revision 3. Methodology from Containment elemental iodine removal iodine by wall deposition (See Section 2.1.2) Atmosphere via deposition on wetted on wetted surfaces inside surfaces. containment is credited in accordance with SRP 6.5.2. Ii. Control Room Inleakage II-1 Control Room Inleakage 350 cfm unfiltered. Same as Revision 2 300 cfm unfiltered. Input Following CREV System (See Section 2.5.2) Initiation. Ill. Containment Leakage Pathway 111-1 Containment Leakage Rate Containment leak rate is Same as Revision 2 The leakage rate is not Input Reduction reduced by 50% after 24 reduced at 24 hours (or (See Section 2.2) hours. subsequently during the LOCA duration) in response to a reduction in containment pressure. IV. ESF Leakage Pathway IV-I ESF Leakage Rate I gpm 2.85 gpm Same as Revision 3. Input (See Section 2.3)

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO. 4 PAGE NO. 69 of 117 Table 1E Comparison of Modeling Differences between Revision 2, Revision 3, and Revision 4 of Calculation H-I-ZZ-MDC-1880 Item Changed Parameter Calc H-I-ZZ-MDC-1880, Cale H-1-ZZ-MDC- I Calc H-I-ZZ-MDC-1880, Type of Change

     #                                        Revision 2                1880, Revision 3                 Revision 4         Input or Methodology V. MSIV Leakage Pathway V-I         MSIV Leakage Rate         MSIV leak rate is reduced     Same as Revision 2.           The leakage rate is not             Input Reduction                 by 50% after 24 hours,                                      reduced at 24 hours (or       (See Section 2.4) subsequently during the LOCA duration) in response to a reduction in containment pressure.

V-2 MSIV Leakage One failed main steam line One failed main steam Same as Revision 3. Methodology Configuration with maximum TS allowed line with 150 scfli leakage (See Section 2.4) leakage of 150 scflh and 2 and 1 intact main steam intact main steam lines with line with 100 scfh 50 scflh MSIV leakage from leakage. each line. V-3 MSIV Leakage Volumetric MSIV volumetric leakage MSIV leakage flow rates Same as Revision 3. Input Flow Rates flow rates are from the are the technical (See Section 2.4.3) technical specification specification volumetric limits specified at standard flow rate limits (scfm), temperature and pressure adjusted to containment conditions, post LOCA temperature and pressure conditions, using the Ideal Gas Law. V-4 Main Steam Line Flow Plug flow model, credit is Well mixed flow model. Same as Revision 3. Methodology Model taken for holdup and No credit for holdup. (See Section 2.4.2) radiodecay in the failed and intact main steam lines. V-5 Number of Main Steam Failed Main Steam Line Failed Main Steam Line Same as Revision 2 Methodology Pipe Volume Nodes for 1 pipe volume. 2 pipe volumes (between (See Sections 2.4.1 and Aerosol Deposition and inboard and outboard 2.4.4) Elemental Iodine Removal Intact Main Steam Line MSIV and between I pipe volume, outboard MSIV and TSV). Intact Main Steam Line 2 pipe volumes (between RPV and outboard MSIV and between outboard MSIV and TSV).

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO. 4 PAGE NO. 70 of 117 Table 1E Comparison of Modeling Differences between Revision 2, Revision 3, and Revision 4 of Calculation H-1-ZZ-MDC-1880 Item Changed Parameter Calc H-1-ZZ-MDC-1880, Calc H-1-ZZ-MDC- Calc H-1-ZZ-MDC-1880, Type of Change Revision 2 1880, Revision 3 Revision 4 Input or Methodology V-6 Pipe Surface Area for Based on horizontal pipe Based on pipe horizontal Same as Revision 3. Methodology Aerosol Deposition surface area (n x D x L). projected area (D x L). (See Section 2.4.4) Aerosol deposition is Aerosol deposition in credited in failed main failed main steam line is steam line from RPV to credited between inboard inboard MSIV and from MSIV and TSV. Aerosol outboard MSIV to TSV. deposition is credited in Aerosol deposition is intact main steam line credited in intact main from RPV to TSV. steam line from RPV to TSV. V-7 Percentile Settling Velocity 4 0 th % for both failed and 5 0 th % upstream of Same as revision 2. Input for Aerosol Deposition intact main steam lines, outboard MSIV for both (See Section 2.4.4) failed and intact main steam lines. 3 0 th % downstream of outboard MSIV for both failed and intact main steam lines. V-8 Time Credited for Aerosol 0 to 30 days 0 to 96 hours Same as Revision 3. Methodology Deposition and Elemental (See Sections 2.4.1 and Iodine Removal in Main 2.4.4) Steam Piping. V-9 Elemental Iodine Removal Calculated using the J. E. 50% (i.e., Same as Revision 2. Methodology Within Main Steam Lines Cline methodology. decontamination factor of (See Section 2.4.5)

2) due to natural removal of elemental iodine in each main steam line volume per Appendix B of AEB-98-03.

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO. 4 PAGE NO. 71 of 117 Table 1E Comparison of Modeling Differences between Revision 2, Revision 3, and Revision 4 of Calculation H-l-ZZ-MDC-1880 Item Changed Parameter Calc H-1-ZZ-MDC-1880, Cale H-1-ZZ-MDC- Calc H-I-ZZ-MDC-1880, Type of Change

     #                                           Revision 2              1880, Revision 3               Revision 4        Input or Methodology VI. Primary Containment Isolation Valve Release Pathway VI-1         Release of Containment      This pathway not            Reactor coolant activity    Same as Revision 2.            Methodology (i.e., Drywell) Airborne    considered.                 released with drywell                                    (See Section 2.2)

Activity via Primary atmosphere from 0 to 120 Containment Isolation seconds post LOCA via Valves. open primary containment isolation valves (i.e., drywell and suppression chamber purge exhaust isolation valves). Direct release to the environment.

ICALCULATION NO. H-1-ZZ-MDC-1880 IREVISION NO.4 1PAGE NO.72 of 117 Table iF Elemental Iodine Deposition Velocity - MSIV Leakage Time Temp Temp Deposition Deposition Degree* Degree (2809/T -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 1.01OE-04 240 170 349.82 -4.47 0.011446 1.145E-04 480 150 338.71 -4.21 0.014896 1.490E-04 720

• From Design Input 5.3.1.9

ICALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO.4 1PAGE NO.73 of 117 1 Table IG Elemental Iodine Deposition Rate - MSIV Failed Line Volume VII Deposition Main Steam Line Elemental Velocity Total Total Iodine Time Surface Volume Removal Area Rate (m 2 ) 3 Hr m/sec (m ) (hr-) A B C (AxB)x3600/C 0 2.076E-05 203.66 35.17 0.4328 3 2.442E-05 203.66 35.17 0.5091 6 4.630E-05 203.66 35.17 0.9651 24 7.248E-05 203.66 35.17 1.5109 96 1.010E-04 203.66 35.17 2.1047 240 1.145E-04 203.66 35.17 2.3862 480 1.490E-04 203.66 35.17 3.1053 720 A From Table IF B and C From Section 7.3.2.5 Table 11 Elemental Iodine Deposition Rate - MSIV Intact Line Volume V33 Deposition Main Steam Line Elemental Velocity Total Total Iodine Time Surface Volume Removal Area Rate 3 Hr m/sec (mi) (mi ) (hr-) A* B C (AxB)x3600/C 0 2.076E-05 217.69 43.63 0.3730 3 2.442E-05 217.69 43.63 0.4386 6 4.630E-05 217.69 43.63 0.8316 24 7.248E-05 217.69 43.63 1.3018 96 1.01OE-04 217.69 43.63 1.8135 240 1.145E-04 217.69 43.63 2.0560 480 1.490E-04 217.69 43.63 2.6756 720 A From Table IF B and C From Section 7.3.3.7

ICALCULATION NO. H-l-ZZ-MDC-1880 I REVISION NO.4 1PAGE NO.74 of 117 Table 1I Elemental Iodine Resuspension Rate - MSIV Leakage Post-LOCA Temp Temp Resuspension Time Degree Degree -600/T Rate (hr) F K (hr-') 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 Resuspension Rate (sec) 1 = 2.32 (2.00) x 10-5 e"600/T 4.32 x 10-5 e"600/T Resuspension Rate (hr)-1 = 4.32 x 3600 x 10-5 e"600/T Table 1J Net Elemental Iodine Removal Rate - MSIV Failed Line Volume V11 Iodine Iodine Net Iodine Post-LOCA Temp Removal Resuspension Removal Time Degree Rate Rate Rate A B C=A-B (hr) F (hr-1) (hr-1) (hr-1) 0 340 0.4328 0.0403 0.3926 3 320 0.5091 0.0389 0.4701 6 250 0.9651 0.0340 0.9312 24 208 1.5109 0.0309 1.4800 96 180 2.1047 0.0287 2.0759 240 170 2.3862 0.0280 2.3582 480 150 3.1053 0.0265 3.0789 720 _ 1 1 A From Table 1G B From Table 11

I CALCULATION NO. H-1-ZZMDC-1880 I RVISION NO. 4 1 PAGE NO. 75 of 117 Table 1K Net Elemental Iodine Removal Rate - Intact Line Volume V33 Iodine Iodine Net Iodine Post-LOCA Temp Removal Resuspension Removal Time Degree Rate Rate Rate A B C=A-B (hr) F (hr-1) (hr-1) (hr-1) 0 340 0.3730 0.0403 0.3327 3 320 0.4386 0.0389 0.3997 6 250 0.8316 0.0340 0.7976 24 208 1.3018 0.0309 1.2709 96 180 1.8135 0.0287 1.7847 240 170 2.0560 0.0280 2.0280 480 150 2.6756 0.0265 2.6492 720 A From Table 1H B From Table 11

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO. 4 PAGE NO. 76 of 117 Table IL Elemental Iodine Removal Efficiency - MSIV Failed Line Volume V11 Net Elemental Post-LOCA Temp Iodine Iodine Time Degree Removal Removal Rate Efficiency A B (hr) F (hr-1) (%) 0 340 0.3926 32.47 3 320 0.4701 37.51 6 250 0.9312 60.59 24 208 1.4800 77.24 96 180 2.0759 87.46 240 170 2.3582 90.54 480 150 3.0789 95.40 720 A From Table 1J B = l-e^-A Table 1M Elemental Iodine Removal Efficiency - Intact Line Volume V33 Net Elemental Post-LOCA Temp Iodine Iodine Time Degree Removal Removal Rate Efficiency A B (hr) F (hr-i) (%) 0 340 0.3327 28.30 3 320 0.3997 32.95 6 250 0.7976 54.96 24 208 1.2709 71.94 96 180 1.7847 83.22 240 170 2.0280 86.84 480 150 2.6492 92.93 720 A From Table IK B = 1-e^-A

CALCULATION NO. H-1-ZZ-MDC-1880 I REVISION NO. 4 PAGE NO.77 of 117 Table 2 Hope Creek Main Steam Horizontal Piping Volume and Projected Horizontal Pipe Inside Surface Area Main Steam Piping Main Steam Piping Horizontal Pipe Projected Horizontal Pipe Inside Volume Inside Surface Area (ft3 ) (ft 2) Header ] Header Header Header A J D A D Piping Between RPV Nozzle and Inboard MSIV N/A ] 76.83 j N/A 49.89 Piping Between Inboard and Outboard MSIVs 101.89 ] 101.89 66.18 66.18 Piping Between Outboard MSIV and Turbine' Stop valve 1065.25 ] 1062.21 631.62 629.82 Total 1167.14 1240.93 697.8 745.89 Table 3 Hope Creek Main Steam Piping Volume Main Steam Piping Inside Volume (ft 3) Header A ] Header D MSIV Failed Line Intact Line Piping Between RPV Nozzle and Inboard MSIV N/A 301.81 Piping Between Inboard and Outboard MSIVs 101.89 101.89 Piping Between Outboard MSIV and Turbine Stop valve 1139.04 1136.00 Total Inside Volume 1,240.93 1,539.70

ICALCULATION NO. H-1-ZZ-MDC-1880 IREVISION NO.4 PAGE NO.78 of 117 Table 4 Main Steam Piping Parameters Used In MSIV Leakage Release Path Model Main Steam Piping Between RPV Nozzle and Outboard Main Steam MSIV Header Horizontal ID Inside Surface Volume Area ________________ (ft3) (ft 2) MSIV Failed Line Steam Header A* 1167.14 697.8 (First Shortest Line) MSIV Intact Line Steam Header D 1240.93 745.89 (Second Shortest Line) Main Steam Header Volume and Surface Area Information From Table 2

                           *Pipe Segment Between RPV and Inboard MSIV Excluded Table 5 Rate Constant for MSIV Leakage Release Path with                4 0 th%/ 4 0 th%    Settling Velocities Settling        Horizontal         Horizontal              Rate Velocity           Settling               Pipe          Constant for Steam Header                     9St              Area             Volume             Settling X, 2                    3 (ft/hr)            (ft )                (ft )            (hr-')

A B C D MSIV Failed Line - Header A Inboard MSIV To 9.56 697.8 1167.14 5.72 Turbine Stop Valve SV-3 MSIV Intact Line - Header B RPV Nozzle To 9.56 745.89 1240.93 5.75 Turbine Stop Valve I I I A = 4 0 th Percentile Settling Velocity = 0.00081 m/sec x 3.28 ft/m x 3600 sec/hr = 9.56 ft/hr for main steam lines upstream of outboard MSIVs B and C From Table 4 and D = X.s = (Ax B)/C

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO. 4 1PAGE NO.79 of 117 Table 6 Gravitational Deposition Aerosol Removal Efficiency On Horizontal Pipe Surface With 4 0 th%/ 4 0 th% Settling Velocity (250 scfh) Post- Post-LOCA Settling Well Volumetric Aerosol LOCA Settling Well Volumetric Aerosol Time Rate Mixed Flow Removal Time Rate Mixed Flow Removal Interval Constant Volume Rate Efficiency Interval Constant Volume Rate Efficiency

                 %s          V2                     MSIV Failed                         ?"              V4                       Intact A           B                           Line                           A                B                        Line (hr)       (hr-')       (ft3)       (ft3/hr)         (%)            (hr)          (hr-')           (ft 3)     (ft3/hr)         (%)

MSIV Failed Main Steam Line Between Inboard MSIV and TSV Intact Main Steam Line Between RPV and TSV 0-24 5.72 1240.93 150.00 97.93 0-24 5.75 1539.70 100.00 98.88 24-96 5.72 1240.93 150.00 97.93 24-96 5.75 1539.70 100.00 98.88 96-720 5.72 1240.93 150.00 0.00 96-720 5.75 1539.70 100.00 0.00 A From Table 5 B From Table 3 Table 7 MSIV Leak Rate In Different Control Volume (Total = 250 scfh and Max = 150 scfh) Post-LOCA MSIV Leak Rate In Various Control Volumes (cfh)/(cfm) Time Drywell To Volume V1 Drywell To Volume V33 Interval MSIV Failed To Intact Line To (hr) Volume V1, Atmosphere Volume V33 Atmosphere 48.47 150.00 32.31 100.00 0-2 0.808 2.500 0.539 1.667 26.77 150.00 17.84 100.00 2-720 0.446 2.500 0.297 1.667 MSIV Leak Rate Information From Section 7.2

CALCULATION NO. H-1-ZZ-MDC-1880 I VISION NO. 4 1PAGE NO.80 of 117 1 Table 8 Conversion of Iodine Activity Into Iodine Atom Drywell Region @ 0.5 hr Iodine Isotopic Isotope Activity Atoms (Atoms Per Iodine (Curie) Curie) Fraction A B Ci = Bi/Ai Di = Bi/EB 1-131 4.247E+06 1.575E+23 3.708E+16 7.589E-01 1-132 5.673E+06 2.507E+21 4.420E+14 1.208E-02 1-133 8.630E+06 3.449E+22 3.997E+15 1.662E-01 1-134 6.499E+06 1.095E+21 1.685E+14 5.276E-03 1-135 9.400E+06 1.194E+22 1.270E+15 5.753E-02 Total 2.075E+23 1.000E+00 A and B From RADTRAD Run CUTOFF.oO output file @ 0.5 hr from Containment Compartment Nuclide Inventory Table 9 Elemental Iodine Activity @ DF of 200 Iodine Core Iodine Elemental Iodine Isotope Core Thermal Core Iodine Activity (Atoms Iodine Inventory Power Activity Released In Per Atoms Level Drywell Curie) (Ci/MWt) MWt (Ci) (Ci) A B C=AxB D=CxO.3x0.0485 E F=DxE 1-131 2.670E+04 3917 1.046E+08 1.522E+06 3.708E+16 5.643E+22 1-132 3.870E+04 3917 1.516E+08 2.206E+06 4.420E+14 9.749E+20 1-133 5.510E+04 3917 2.158E+08 3.140E+06 3.997E+15 1.255E+22 1-134 6.060E+04 3917 2.374E+08 3.454E+06 1.685E+14 5.818E+20 1-135 6.220E+04 3917 2.436E+08 3.545E+06 1.270E+15 4.503E+21 Total Elemental Iodine Atoms 7.504E+22 Total Iodine Elemental Atoms (@DF of 200 3.752E+20 A from Table ID B From Section 7.9 D Reference 10.1, Appendix A, Section 3.3 E From table 8

I CALCULATION NO. H-1-ZZ-MDC-1880 I REVISION NO.4 I PAGE NO. 81 of 117 Table 10 Post-LOCA Cont. Leakage Iodine Activity Deposited on CR Charcoal Filter Iodine Fraction Elemental and Iodine Iodine Isotope Atoms Per Of Iodine Organic Iodine Atoms on Activity Curie Atoms On CR Charcoal CR Charcoal CR Charcoal Filter Filter 720 Hrs At 720 Hrs At 720 Hrs Ci A B C Di =Bi

• C Ei = Di / Ai 1-131 3.708E+16 7.589E-01 3.107E+17 2.358E+17 6.359E+00 1-132 4.420E+14 1.208E-02 3.754E+15 8.494E+00 1-133 3.997E+15 1.662E-01 5.165E+16 1.292E+01 1-134 1.685E+14 5.276E-03 1.639E+15 9.731E+00 1-135 1.270E+15 5.753E-02 1.788E+16 1.407E+01 Total Iodine Atoms/Activity 3.107E+ 17 5.158E+01 A and B From Table 8 C From Section 7.6.3A

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO. 4 PAGE NO.82 of 117 Table 11 Relationship of Aerosol Mass and Activity CR Region @ 2.0 hr Aerosol Isotopic Isotope Activity Mass Mass Per Ci Aerosol (Curie) (kg) (kg/Ci) Fraction A Bi Ci = Bi/Ai Di = Bi/ZB Co-58 1.120E-08 3.523E-16 3.145E-08 4.245E-07 Co-60 1.342E-08 1.187E-14 8.846E-07 1.430E-05 Rb-86 6.277E-07 7.714E-15 1.229E-08 9.295E-06 Sr-89 1.576E-05 5.425E-13 3.442E-08 6.537E-04 Sr-90 1.549E-06 1.135E-11 7.331E-06 1.368E-02 Sr-91 2.687E-05 7.411 E- 15 2.759E- 10 8.930E-06 Sr-92 1.270E-05 1.010E-15 7.956E-1 1 1.217E-06 Y-90 3.381E-08 6.214E-17 1.838E-09 7.488E-08 Y-91 2.077E-07 8.471E-1 5 4.078E-08 1.021E-05 Y-92 2.901E-06 3.015E-16 1.039E-10 3.633E-07 Y-93 2.127E-07 6.376E-17 2.997E-10 7.682E-08 Zr-95 2.842E-07 1.323E-14 4.655E-08 1.594E-05 Zr-97 7.932E-07 4.149E-16 5.231E-10 5.OOOE-07 Nb-95 2.856E-07 7.305E-15 2.557E-08 8.802E-06 Mo-99 3.662E-06 7.634E-15 2.085E-09 9.199E-06 Tc-99m 3.271E-06 6.220E- 16 1.902E- 10 7.495E-07 Ru-103 5.637E-06 1.747E-13 3.099E-08 2.105E-04 Ru-105 1.449E-06 2.155E-16 1.488E-10 2.597E-07 Ru-106 2.155E-06 6.442E- 13 2.989E-07 7.762E-04 Rh-105 1.849E-06 2.191E-15 1.1 85E-09 2.640E-06 Sb-127 4.045E-06 1.515E-14 3.745E-09 1.825E-05 Sb-129 9.032E-06 1.606E-15 1.778E-10 1.935E-06 Te-127 4.051E-06 1.535E-15 3.789E-10 1.849E-06 Te-127m 5.441E-07 5.769E- 14 1.060E-07 6.951E-05 Te- 129 1.026E-05 4.899E-16 4.775E-11 5.903E-07 Te-129m 1.819E-06 6.039E-14 3.319E-08 7.276E-05

I CALCULATION NO. H--ZZ-MDC-1880 I REVSION NO. 4 1 PAGE NO.83 of 117 Table 11 (Cont'd) Relationship of Aerosol Mass and Activity CR Region @ 2.0 hr Aerosol Isotopic Isotope Activity Mass Mass Per Ci Aerosol (Curie) (kg) (kg/Ci) Fraction A Bi Ci = Bi/Ai Di = Bi/EB Te-131m 3.870E-05 4.853E-14 1.254E-09 5.848E-05 Te-132 5.489E-05 1.808E-13 3.294E-09 2.178E-04 Cs-134 5.304E-05 4.100E-11 7.729E-07 4.940E-02 Cs-136 1.836E-05 2.505E-13 1.364E-08 3.019E-04 Cs-137 6.703E-05 7.706E-10 1.150E-05 9.285E-01 Ba-139 1.062E-05 6.492E- 16 6.114E-I 1 7.823E-07 Ba-140 2.791E-05 3.812E-13 1.366E-08 4.594E-04 La-140 7.932E-07 1.427E-15 1.799E-09 1.720E-06 La-141 1.859E-07 3.287E-17 1.768E-10 '3.961E-08 La-142 1.043E-07 7.286E-18 6.986E-I1 8.779E-09 Ce-141 6.652E-07 2.335E-14 3.510E-08 2.813E-05 Ce-143 5.933E-07 8.935E-16 1.506E-09 1.077E-06 Ce-144 1.088E-06 3.411E-13 3.135E-07 4.11OE-04 Pr-143 2.401E-07 3.566E-15 1.485E-08 4.297E-06 Nd-147 1.056E-07 1.305E-15 1.236E-08 1.573E-06 Np-239 7.469E-06 3.220E- 14 4.311E-09 3.879E-05 Pu-238 1.326E-09 7.743E- 14 5.841 E-05 9.330E-05 Pu-239 1.599E-10 2.572E- 12 1.609E-02 3.099E-03 Pu-240 2.068E-10 9.074E-13 4.388E-03 1.093E-03 Pu-241 5.997E-08 5.822E-13 9.708E-06 7.015E-04 Am-241 2.700E- 11 7.866E- 15 2.914E-04 9.478E-06 Cm-242 6.391E-09 1.928E-15 3.017E-07 2.324E-06 Cm-244 3.073E-10 3.799E- 15 1.236E-05 4.578E-06 Total 8.299E- 10 1.00E+00 A and B From RADTRAD Run HEPU300CLOO output file @ 2.0 hr from Control Room Compartment Nuclide Inventory

CALCULATION NO. H-1-ZZ-MDC-1880 IREVISION NO. 4 PAGE NO. 84 of 117 Table 12 Post-LOCA Total Aerosol Isotopic Activity On CR HEPA Filter @ 720 Hrs Aerosol Fraction Total Aerosol Isotopic Isotope Mass Per Ci of CR Filter Aerosol Mass Aerosol Activity Aerosol Aerosol Mass On CR Filter On CR Filter At 720 Hr At 720 Hr At 720 Hr (kg/Ci) (kg) (kg) (Ci) A Bi C Di =Bi

• C Ei = Di / Ai Co-58 3.145E-08 4.245E-07 1.350E-07 5.730E-14 1.822E-06 Co-60 8.846E-07 1.430E-05 1.350E-07 1.931E-12 2.183E-06 Rb-86 1.229E-08 9.295E-06 1.350E-07 1.255E-12 1.021E-04 Sr-89 3.442E-08 6.537E-04 1.350E-07 8.824E- 11 2.564E-03 Sr-90 7.331E-06 1.368E-02 1.350E-07 1.847E-09 2.519E-04 Sr-91 2.759E-10 8.930E-06 1.350E-07 1.206E-12 4.370E-03 Sr-92 7.956E- 11 1.217E-06 1.350E-07 1.643E-13 2.065E-03 Y-90 1.838E-09 7.488E-08 1.350E-07 1.01IE-14 5.500E-06 Y-91 4.078E-08 1.021E-05 1.350E-07 1.378E-12 3.379E-05 Y-92 1.039E-10 3.633E-07 1.350E-07 4.905E-14 4.720E-04 Y-93 2.997E-10 7.682E-08 1.350E-07 1.037E-14 3.460E-05 Zr-95 4.655E-08 1.594E-05 1.350E-07 2.152E-12 4.623E-05 Zr-97 5.231E-10 5.OOOE-07 1.350E-07 6.750E-14 1.290E-04 Nb-95 2.557E-08 8.802E-06 1.350E-07 1.188E-12 4.647E-05 Mo-99 2.085E-09 9.199E-06 1.350E-07 1.242E-12 5.956E-04 Tc-99m 1.902E-10 7.495E-07 1.350E-07 1.012E-13 5.320E-04 Ru-103 3.099E-08 2.105E-04 1.350E-07 2.841E-11 9.170E-04 Ru-105 1.488E-10 2.597E-07 1.350E-07 3.506E-14 2.357E-04 Ru-106 2.989E-07 7.762E-04 1.350E-07 1.048E-10 3.506E-04 Rh-105 1.185E-09 2.640E-06 1.350E-07 3.563E-13 3.008E-04 Sb-127 3.745E-09 1.825E-05 1.350E-07 2.464E-12 6.579E-04 Sb-129 1.778E-10 1.935E-06 1.350E-07 2.613E-13 1.469E-03 Te- 127 3.789E-10 1.849E-06 1.350E-07 2.497E- 13 6.589E-04 Te-127m 1.060E-07 6.951E-05 1.350E-07 9.384E-12 8.851E-05 Te-129 4.775E-1I 5.903E-07 1.350E-07 7.969E-14 1.669E-03 Te-129m 3.319E-08 7.276E-05 I 1.350E-07 9.823E-12 2.959E-04

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO.4 PAGE NO.85 of 117 1 Table 12 (Cont'd) Post-LOCA Total Aerosol Isotopic Activity On CR HEPA Filter @ 720 Hrs Aerosol Fraction Total Aerosol Isotopic Isotope Mass Per Ci of CR Filter Aerosol Mass Aerosol Activity Aerosol Aerosol Mass On CR Filter On CR Filter At 720 Hr At 720 Hr At 720 Hr (kg/Ci) (kg) (kg) (Ci) A Bi C Di =Bi

• C Ei = Di / Ai Te-131m 1.254E-09 5.848E-05 1.350E-07 7.895E-12 6.295E-03 Te-132 3.294E-09 2.178E-04 1.350E-07 2.941 E- 11 8.928E-03 Cs-134 7.729E-07 4.940E-02 1.350E-07 6.669E-09 8.628E-03 Cs-136 1.364E-08 3.019E-04 1.350E-07 4.075E- 11 2.987E-03 Cs-137 1.150E-05 9.285E-01 1.350E-07 1.253E-07 1.090E-02 Ba-139 6.114E-11 7.823E-07 1.350E-07 1.056E-13 1.727E-03 Ba-140 1.366E-08 4.594E-04 1.350E-07 6.202E- 11 4.540E-03 La-140 1.799E-09 1.720E-06 1.350E-07 2.321E-13 1.290E-04 La-141 1.768E-10 3.961E-08 1.350E-07 5.347E-15 3.024E-05 La-142 6.986E- I1 8.779E-09 1.350E-07 1.185E-15 1.697E-05 Ce-141 3.51 OE-08 2.813E-05 1.350E-07 3.798E-12 1.082E-04 Ce- 143 1.506E-09 1.077E-06 1.350E-07 1.453E-13 9.652E-05 Ce- 144 3.135E-07 4.11OE-04 1.350E-07 5.548E- 11 1.770E-04 Pr-143 1.485E-08 4.297E-06 1.350E-07 5.801E-13 3.906E-05 Nd-147 1.236E-08 1.573E-06 1.350E-07 2.124E-13 1.718E-05 Np-239 4.311E-09 3.879E-05 1.350E-07 5.237E-12 1.215E-03 Pu-238 5.841 E-05 9.330E-05 1.350E-07 1.260E- 11 2.156E-07 Pu-239 1.609E-02 3.099E-03 1.350E-07 4.184E-10 2.601E-08 Pu-240 4.388E-03 1.093E-03 1.350E-07 1.476E-10 3.363E-08 Pu-241 9.708E-06 7.015E-04 1.350E-07 9.471E- 11 9.756E-06 Am-241 2.914E-04 9.478E-06 1.350E-07 1.280E-12 4.392E-09 Cm-242 3.017E-07 2.324E-06 1.350E-07 3.137E-13 1.040E-06 Cm-244 1.236E-05 4.578E-06 1.350E-07 6.180E-13 5.OOOE-08 Total CR Aerosol Mass/Activity @ 720 hrs 1.350E-07 6.375E-02 A and B From Table 11 C From Section 7.6.3A

CALCULATION NO. H-1-ZZ-MDC-1880 I REVISION NO. 4 1PAGE NO.86 of 117 Table 13 720-hrs Post-LOCA Total Iodine and Aerosol Activity On CREF Charcoal/HEPA Filters (Ci) Total Total Isotope Activity Isotope Activity Ci Ci A A Co-58 1.822E-06 Te-131m 6.295E-03 Co-60 2.183E-06 Te-132 8.928E-03 Rb-86 1.021E-04 Cs-134 8.628E-03 Sr-89 2.564E-03 Cs-136 2.987E-03 Sr-90 2.519E-04 Cs-137 1.090E-02 Sr-91 4.370E-03 Ba-139 1.727E-03 Sr-92 2.065E-03 Ba-140 4.540E-03 Y-90 5.500E-06 La-140 1.290E-04 Y-91 3.379E-05 La-141 3.024E-05 Y-92 4.720E-04 La- 142 1.697E-05 Y-93 3.460E-05 Ce-141 1.082E-04 Zr-95 4.623E-05 Ce- 143 9.652E-05 Zr-97 1.290E-04 Ce-I144 1.770E-04 Nb-95 4.647E-05 Pr-143 3.906E-05 Mo-99 5.956E-04 Nd-147 1.718E-05 Tc-99m 5.320E-04 Np-239 1.215E-03 Ru-103 9.170E-04 Pu-238 2.156E-07 Ru-105 2.357E-04 Pu-239 2.601E-08 Ru-106 3.506E-04 Pu-240 3.363E-08 Rh-105 3.008E-04 Pu-241 9.756E-06 Sb-127 6.579E-04 Am-241 4.392E-09 Sb- 129 1.469E-03 Cm-242 1.040E-06 Te-127 6.589E-04 Cm-244 5.OOOE-08 Te-127m 8.851E-05 1-131 6.359E+00 Te-129 1.669E-03 1-132 8.494E+00 Te-129m 2.959E-04 1-133 1.292E+01 1-134 9.731E+00 1-135 1.407E+01 A - Aerosol From Table 12 A - Iodine From Table 10

CALCULATION NO. H-1-ZZ-MDC-1880 I REVISION NO. 4 IPAGE NO.87 of 117 Table 14 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 8.161E-06 O.OOOE+00 1.727E-05 2.543E-05 Co-60 9.771E-06 O.OOOE+00 2.068E-05 3.045E-05 Kr-85 1.836E+00 O.OOOE+00 1.671E+00 3.507E+00 Kr-85m 3.592E+O1 O.OOOE+00 3.237E+01 6.828E+01 Kr-87 5.160E+O I O.OOOE+00 4.514E+01 9.673E+01 Kr-88 9.123E+01 O.OOOE+00 8.171E+01 1.729E+02 Rb-86 2.919E-02 O.OOOE+00 3.132E-03 3.232E-02 Sr-89 1.148E-02 O.OOOE+00 2.431E-02 3.579E-02 Sr-90 1.128E-03 O.OOOE+00 2.387E-03 3.514E-03 Sr-91 2.118E-02 O.OOOE+00 4.481E-02 6.598E-02 Sr-92 1.221E-02 O.OOOE+00 2.580E-02 3.801E-02 Y-90 1.597E-05 O.OOOE+00 3.403E-05 5.001E-05 Y-91 1.484E-04 O.OOOE+00 3.141E-04 4.625E-04 Y-92 9.437E-04 O.OOOE+00 2.041E-03 2.985E-03 Y-93 1.669E-04 O.OOOE+00 3.531E-04 5.200E-04 Zr-95 2.071E-04 O.OOOE+00 4.383E-04 6.454E-04 Zr-97 6.039E-04 O.OOOE+00 1.278E-03 1.882E-03 Nb-95 2.080E-04 O.OOOE+00 4.403E-04 6.483E-04 Mo-99 2.697E-03 O.OOOE+00 5.708E-03 8.405E-03 Tc-99m 2.383E-03 O.OOOE+00 5.044E-03 7.427E-03 Ru- 103 4.108E-03 O.OOOE+00 8.696E-03 1.280E-02 Ru-105 1.250E-03 O.OOOE+00 2.643E-03 3.893E-03 Ru-106 1.570E-03 O.OOOE+00 3.322E-03 4.892E-03 Rh-105 1.351E-03 O.OOOE+00 2.859E-03 4.209E-03 Sb-127 2.970E-03 O.OOOE+00 6.285E-03 9.254E-03 Sb-129 7.829E-03 O.OOOE+00 1.656E-02 2.439E-02 Te-127 2.960E-03 O.OOOE+00 6.265E-03 9.226E-03 Te-127m 3.962E-04 O.OOOE+00 8.386E-04 1.235E-03 Te- 129 8.327E-03 O.OOOE+00 1.763E-02 2.595E-02 Te-129m 1.325E-03 O.OOOE+00 2.803E-03 4.128E-03

CALCULATION NO. H-1-ZZ-MDC-1880 I REVISION NO. 4 PAGE NO. 88 of 117 Table 14 (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 (Ci) A B C A+B+C Te-131m 2.890E-02 O.OOOE+00 6.116E-02 9.005E-02 Te-132 4.036E-02 O.OOOE+00 8.541E-02 1.258E-01 1-131 1.245E+O1 3.802E+00 3.OOOE+00 1.926E+01 1-132 1.741E+01 5.064E+00 3.715E+00 2.619E+01 1-133 2.559E+01 7.752E+00 6.043E+00 3.938E+01 1-134 2.547E+01 6.562E+00 3.698E+00 3.573E+01 1-135 2.860E+01 8.511E+00 6.447E+00 4.356E+O1 Xe-133 2.918E+02 8.516E-02 2.656E+02 5.575E+02 Xe-135 1.061E+02 1.104E+00 9.739E+01 2.045E+02 Cs-134 2.460E+00 O.OOOE+00 2.642E-01 2.724E+00 Cs-136 8.550E-01 O.OOOE+00 9.169E-02 9.467E-01 Cs-137 3.108E+00 O.OOOE+00 3.338E-01 3.442E+00 Ba-139 1.336E-02 O.OOOE+00 2.821 E-02 4.157E-02 Ba-140 2.038E-02 O.OOOE+00 4.312E-02 6.350E-02 La-140 3.309E-04 O.OOOE+00 7.069E-04 1.038E-03 La- 141 1.640E-04 O.OOOE+00 3.467E-04 5.107E-04 La-142 1.238E-04 O.OOOE+00 2.615E-04 3.853E-04 Ce-141 4.847E-04 O.OOOE+00 1.026E-03 1.511E-03 Ce-143 4.421E-04 O.OOOE+00 9.355E-04 1.378E-03 Ce-144 7.923E-04 O.OOOE+00 1.677E-03 2.469E-03 Pr-143 1.745E-04 O.OOOE+00 3.693E-04 5.437E-04 Nd-147 7.713E-05 O.OOOE+00 1.632E-04 2.404E-04 Np-239 5.512E-03 O.OOOE+00 1.167E-02 1.718E-02 Pu-238 9.653E-07 O.OOOE+00 2.043E-06 3.008E-06 Pu-239 1.164E-07 O.OOOE+00 2.464E-07 3.628E-07 Pu-240 1.506E-07 O.OOOE+00 3.187E-07 4.692E-07 Pu-241 4.367E-05 O.OOOE+00 9.244E-05 1.361E-04 Am-241 1.965E-08 O.OOOE+00 4.160E-08 6.125E-08 Cm-242 4.655E-06 O.OOOE+00 9.852E-06 1.451E-05 Cm-244 2.238E-07 O.OOOE+00 4.737E-07 6.975E-07 A From HEPU300CLOO.oO B From HEPU300ESOO.oO C From HIN300MSOO.oO

ICALCULATION NO. H-°-ZZ-MDC-1880 I REVISION NO.4 PAGE NO.89 of 117 Table 15 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 1.475E-04 0.OOOE+00 4.377E-04 5.851E-04 Co-60 1.766E-04 0.OOOE+00 5.242E-04 7.008E-04 Kr-85 2.540E+01 0.OOOE+00 2.549E+01 5.089E+01 Kr-85m 4.369E+02 0.OOOE+00 4.377E+02 8.747E+02 Kr-87 4.569E+02 0.OOOE+00 4.544E+02 9.113E+02 Kr-88 1.031E+03 0.OOOE+00 1.032E+03 2.063E+03 Rb-86 3.603E-02 O.OOOE+00 2.661E-02 6.264E-02 Sr-89 2.075E-01 0.OOOE+00 6.159E-01 8.234E-01 Sr-90 2.038E-02 0.OOOE+00 6.050E-02 8.088E-02 Sr-91 3.635E-01 0.OOOE+00 1.076E+00 1.439E+00 Sr-92 1.846E-01 0.OOOE+00 5.420E-01 7.266E-01 Y-90 3.812E-04 0.OOOE+00 1.182E-03 1.563E-03 Y-91 2.713E-03 0.OOOE+00 8.071E-03 1.078E-02 Y-92 3.015E-02 0.OOOE+00 9.727E-02 1.274E-01 Y-93 2.874E-03 0.OOOE+00 8.506E-03 1.138E-02 Zr-95 3.742E-03 0.OOOE+00 1.11IE-02 1.485E-02 Zr-97 1.060E-02 0.OOOE+00 3.142E-02 4.203E-02 Nb-95 3.760E-03 0.OOOE+00 1.116E-02 1.492E-02 Mo-99 4.838E-02 0.OOOE+00 1.436E-01 1.919E-01 Tc-99m 4.306E-02 0.OOOE+00 1.278E-01 1.709E-01 Ru-103 7.422E-02 0.OOOE+00 2.203E-01 2.945E-01 Ru-105 2.024E-02 0.OOOE+00 5.971E-02 7.995E-02 Ru-106 2.837E-02 O.OOOE+00 8.421E-02 1.126E-01 Rh-105 2.437E-02 0.OOOE+00 7.233E-02 9.669E-02 Sb-127 5.339E-02 0.OOOE+00 1.584E-01 2.118E-01 Sb-129 1.264E-01 0.OOOE+00 3.728E-01 4.993E-01 Te-127 5.338E-02 0.OOOE+00 1.584E-01 2.118E-01 Te-127m 7.162E-03 0.OOOE+00 2.126E-02 2.842E-02 Te-129 1.405E-01 0.OOOE+00 4.156E-01 5.561E-01 Te-129m 2.394E-02 0.OOOE+00 7.107E-02 9.502E-02

ICALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO.4 1PAGE NO.90 of 117 Table 15 (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-131m 5.138E-01 0.OOOE+00 1.524E+00 2.038E+00 Te-132 7.248E-01 0.OOOE+00 2.15]E+00 2.876E+00 1-131 1.687E+01 1.674E+01 2.287E+01 5.648E+01 1-132 2.248E+0 I 1.879E+01 2.578E+01 6.705E+01 1-133 3.426E+01 3.313E+01 4.501E+01 1.124E+02 1-134 2.830E+01 1.467E+01 1.629E+01 5.926E+0 1 1-135 3.731E+O1 3.396E+01 4.555E+01 1.168E+02 Xe-133 4.027E+03 1.443E+00 4.044E+03 8.073E+03 Xe-135 1.487E+03 1.763E+01 1.507E+03 3.011E+03 Cs-134 3.037E+00 O.OOOE+00 2.247E+00 5.285E+00 Cs-136 1.055E+00 0.OOOE+00 7.787E-01 1.834E+00 Cs-137 3.838E+00 0.OOOE+00 2.840E+00 6.678E+00 Ba-139 1.710E-01 0.OOOE+00 4.968E-01 6.678E-01 Ba-140 3.677E-01 0.OOOE+00 1.091E+00 1.459E+00 La-140 8.622E-03 0.OOOE+00 2.702E-02 3.564E-02 La-141 2.619E-03 O.OOOE+00 7.717E-03 1.034E-02 La-142 1.642E-03 0.OOOE+00 4.782E-03 6.423E-03 Ce-141 8.758E-03 0.OOOE+00 2.600E-02 3.475E-02 Ce-143 7.872E-03 0.OOOE+00 2.335E-02 3.122E-02 Ce-144 1.432E-02 0.OOOE+00 4.250E-02 5.682E-02 Pr-143 3.158E-03 0.OOOE+00 9.376E-03 1.253E-02 Nd-147 1.391E-03 0.OOOE+00 4.130E-03 5.521E-03 Np-239 9.876E-02 0.000E+00 2.930E-01 3.918E-01 Pu-238 1.745E-05 0.OOOE+00 5.179E-05 6.924E-05 Pu-239 2.104E-06 0.OOOE+00 6.246E-06 8.351E-06 Pu-240 2.721E-06 0.OOOE+00 8.078E-06 1.080E-05 Pu-241 7.894E-04 O.OOOE+00 2.343E-03 3.133E-03 Am-241 3.553E-07 0.OOOE+00 1.055E-06 1.410E-06 Cm-242 8.413E-05 0.OOOE+00 2.497E-04 3.338E-04 Cm-244 4.045E-06 0.OOOE+00 I1.201E-05 1.605E-05 A From HEPU300CL0O.oO B From HEPU300ESOO.oO C From HIN300MSOO.oO

ICALCULATION NO. H-1-ZZMDC-1880 REVISION NO.4 PAGE NO.91 of 117 Table 16 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 3.773E-04 O.OOOE+00 2.599E-03 2.976E-03 Co-60 4.520E-04 O.OOOE+00 3.114E-03 3.566E-03 Kr-85 1.341E+02 O.OOOE+00 1.725E+02 3.065E+02 Kr-85m 1.904E+03 O.OOOE+00 2.432E+03 4.336E+03 Kr-87 1.307E+03 O.OOOE+00 1.624E+03 2.931E+03 Kr-88 4.049E+03 O.OOOE+00 5.144E+03 9.193E+03 Rb-86 4.531E-02 O.OOOE+00 1.244E-01 1.698E-01 Sr-89 5.309E-01 O.OOOE+00 3.656E+00 4.187E+00 Sr-90 5.217E-02 O.OOOE+00 3.594E-01 4.116E-01 Sr-91 8.823E-01 O.OOOE+00 5.870E+00 6.752E+00 Sr-92 3.962E-01 O.OOOE+00 2.416E+00 2.812E+00 Y-90 1.287E-03 O.OOOE+00 1.084E-02 1.212E-02 Y-91 7.043E-03 O.OOOE+00 4.916E-02 5.620E-02 Y-92 1.047E-01 O.OOOE+00 9.033E-01 1.008E+00 Y-93 6.996E-03 O.OOOE+00 4.664E-02 5.363E-02 Zr-95 9.574E-03 O.OOOE+00 6.595E-02 7.552E-02 Zr-97 2.633E-02 O.OOOE+00 1.779E-01 2.042E-01 Nb-95 9.624E-03 O.OOOE+00 6.630E-02 7.592E-02 Mo-99 1.229E-0I O.OOOE+00 8.424E-01 9.653E-01 Tc-99m 1.101E-01 O.OOOE+00 7.580E-01 8.681E-01 Ru-103 1.899E-01 O.OOOE+00 1.308E+00 1.498E+00 Ru-105 4.638E-02 O.OOOE+00 2.965E-01 3.429E-01 Ru-106 7.261E-02 O.OOOE+00 5.002E-01 5.728E-01 Rh-105 6.215E-02 O.OOOE+00 4.271E-01 4.892E-01 Sb-127 1.359E-01 O.OOOE+00 9.329E-01 1.069E+00 Sb-129 2.888E-01 O.OOOE+00 1.843E+00 2.131E+00 Te-127 1.363E-01 O.OOOE+00 9.374E-01 1.074E+00 Te-127m 1.833E-02 O.OOOE+00 1.263E-01 1.446E-01 Te-129 3.318E-01 O.OOOE+00 2.165E+00 2.497E+00 Te-129m 6.129E-02 0.000E+00 4.223E-01 4.835E-01

CALCULATION NO. H-I-ZZ-MDC-1880 I REVISION NO. 4 1PAGE NO.92 of 117 Table 16 (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- 131 m 1.293E+00 0.OOOE+00 8.809E+00 1.010E+01 Te-132 1.843E+00 0.OOOE+00 1.265E+01 1.449E+i01 1-131 2.386E+01 6.742E+01 9.854E+01 1.898E+02 1-132 2.907E+01 5.688E+01 9.145E+01 1.774E+02 1-133 4.743E+01 1.280E+02 1.871E+02 3.625E+02 1-134 3.005E+01 2.503E+01 3.317E+01 8.825E+01 1-135 4.940E+01 1.194E+02 1.745E+02 3.433E+02 Xe-133 2.115E+04 1.294E+01 2.722E+04 4.838E+04 Xe-135 7.566E+03 1.425E+02 9.918E+03 1.763E+04 Cs-134 3.823E+00 O.OOOE+00 1.053E+01 1.435E+01 Cs-136 1.326E+00 0.OOOE+00 3.639E+00 4.965E+00 Cs-137 4.830E+00 O.OOOE+00 1.330E+01 1.813E+01 Ba-139 3.184E-01 0.OOOE+00 1.730E+00 2.049E+00 Ba-140 9.396E-01 0.OOOE+00 6.466E+00 7.405E+00 La-140 3.089E-02 0.OOOE+00 2.686E-01 2.995E-01 La-141 5.918E-03 0.OOOE+00 3.747E-02 4.339E-02 La- 142 3.146E-03 0.OOOE+00 1.752E-02 2.066E-02 Ce- 141 2.241 E-02 0.OOOE+00 1.543E-01 1.767E-01 Ce-143 1.984E-02 0.OOOE+00 1.353E-01 1.551E-01 Ce-144 3.665E-02 0.OOOE+00 2.525E-01 2.891E-01 Pr-143 8.097E-03 0.OOOE+00 5.587E-02 6.397E-02 Nd-147 3.555E-03 0.OOOE+00 2.446E-02 2.801E-02 Np-239 2.505E-01 0.OOOE+00 1.716E+00 1.966E+00 Pu-238 4.466E-05 0.OOOE+00 3.077E-04 3.524E-04 Pu-239 5.387E-06 0.OOOE+00 3.712E-05 4.250E-05 Pu-240 6.966E-06 0.OOOE+00 4.799E-05 5.496E-05 Pu-241 2.021E-03 0.OOOE+00 1.392E-02 1.594E-02 Am-241 9.097E-07 0.OOOE+00 6.268E-06 7.178E-06 Cm-242 2.153E-04 0.OOOE+00 1.483E-03 1.699E-03 Cm-244 1.036E-05 0.OOOE+00 7.134E-05 8.170E-05 A From HEPU300CLOO.oO B From HEPU300ESOO.oO C From HIN300MSOO.oO

ICALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO.4 PAGE NO.93 of 117 Table 17 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 Co-58 4.278E-04 O.OOOE+00 6.484E-03 6.912E-03 Co-60 5.126E-04 O.OOOE+00 7.775E-03 8.288E-03 Kr-85 5.695E+02 O.OOOE+00 7.315E+02 1.301E+03 Kr-85m 5.606E+03 O.OOOE+00 7.210E+03 1.282E+04 Kr-87 2.046E+03 O.OOOE+00 2.590E+03 4.636E+03 Kr-88 9.938E+03 O.OOOE+00 1.277E+04 2.271E+04 Rb-86 4.729E-02 O.OOOE+00 2.950E-01 3.423E-01 Sr-89 6.019E-01 O.OOOE+00 9.119E+00 9.721E+00 Sr-90 5.916E-02 O.OOOE+00 8.974E-01 9.565E-01 Sr-91 9.794E-01 O.OOOE+00 1.290E+01 1.388E+01 Sr-92 4.230E-01 O.OOOE+00 4.101E+00 4.524E+00 Y-90 1.643E-03 O.OOOE+00 4.374E-02 4.538E-02 Y-91 8.040E-03 O.OOOE+00 1.275E-01 1.355E-01 Y-92 1.269E-01 O.OOOE+00 2.680E+00 2.807E+00 Y-93 7.774E-03 O.OOOE+00 1.033E-01 1.110E-01 Zr-95 1.086E-02 O.OOOE+00 1.645E-01 1.754E-01 Zr-97 2.949E-02 O.OOOE+00 4.127E-01 4.422E-0 1 Nb-95 1.091E-02 O.OOOE+00 1.655E-01 1.764E-01 Mo-99 1.389E-01 O.OOOE+00 2.063E+00 2.202E+00 Tc-99m 1.248E-01 O.OOOE+00 1.881E+00 2.006E+00 Ru- 103 2.153E-01 O.OOOE+00 3.261E+00 3.476E+00 Ru-105 5.047E-02 O.OOOE+00 5.735E-01 6.239E-01 Ru-106 8.233E-02 O.OOOE+00 1.249E+00 1.331E+00 Rh-105 7.030E-02 O.OOOE+00 1.049E+00 1.120E+00 Sb-127 1.537E-01 O.OOOE+00 2.297E+00 2.451E+00 Sb-129 3.140E-01 O.OOOE+00 3.542E+00 3.856E+00 Te-127 1.544E-01 O.OOOE+00 2.324E+00 2.478E+00 Te-127m 2.079E-02 O.OOOE+00 3.154E-01 3.362E-01 Te-129 3.640E-01 O.OOOE+00 4.400E+00 4.764E+00 Te-129m 6.950E-02 O.OOOE+00 1.054E+00 1.123E+00

ICALCULATION NO. H-1-ZZ-MDC-1880 I REVISION NO.4 PAGE NO.94 of 117 Table 17 (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 1.456E+00 0.OOOE+00 2.109E+01 2.255E+01 Te-132 2.084E+00 0.OOOE+00 3.106E+01 3.315E+01 1-131 2.929E+01 2.615E+02 2.299E+02 5.207E+02 1-132 3.197E+01 1.197E+02 1.598E+02 3.115E+02 1-133 5.679E+01 4.598E+02 4.135E+02 9.300E+02 1-134 3.023E+01 2.976E+01 3.722E+01 9.721E+01 1-135 5.639E+01 3.602E+02 3.429E+02 7.594E+02 Xe-133 8.866E+04 1.115E+02 1.140E+05 2.028E+05 Xe-135 2.738E+04 9.991 E+02 3.652E+04 6.490E+04 Cs-134 3.991E+00 0.OOOE+00 2.503E+01 2.902E+01 Cs-136 1.384E+00 0.OOOE+00 8.615E+00 9.999E+00 Cs-137 5.043E+00 0.OOOE+00 3.163E+01 3.667E+01 Ba-139 3.296E-01 0.OOOE+00 2.330E+00 2.659E+00 Ba-140 1.065E+00 0.OOOE+00 1.608E+01 1.714E+01 La-140 4.018E-02 0.OOOE+00 1.135E+00 1.1 75E+00 La-141 6.412E-03 0.OOOE+00 7.044E-02 7.685E-02 La-142 3.273E-03 0.OOOE+00 2.455E-02 2.783E-02 Ce-141 2.540E-02 0.OOOE+00 3.848E-01 4.102E-01 Ce-143 2.235E-02 0.OOOE+00 3.252E-01 3.475E-01 Ce- 144 4.156E-02 0.OOOE+00 6.303E-01 6.718E-01 Pr-143 9.189E-03 0.OOOE+00 1.402E-01 1.494E-01 Nd-147 4.027E-03 0.OOOE+00 6.077E-02 6.480E-02 Np-239 2.830E-01 0.OOOE+00 4.189E+00 4.472E+00 Pu-238 5.064E-05 0.OOOE+00 7.682E-04 8.189E-04 Pu-239 6.109E-06 0.OOOE+00 9.269E-05 9.880E-05 Pu-240 7.899E-06 0.OOOE+00 1.198E-04 1.277E-04 Pu-241 2.291E-03 O.OOOE+00 3.476E-02 3.705E-02 Am-241 1.032E-06 0.OOOE+00 1.566E-05 1.669E-05 Cm-242 2.441 E-04 0.OOOE+00 3.702E-03 3.946E-03 Cm-244 1.174E-05 0.OOOE+00I 1.781E-04 1.899E-04 A From HEEPU300CLOO.oO B From HEPU300ESOO.oO C From HIN300MS00.o0

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO. 4 IPAGE NO.95 of 117 Table 18 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 Co-58 4.310E-04 0.OOOE+00 1.525E-02 1.568E-02 Co-60 5.164E-04 0.OOOE+00 1.843E-02 1.894E-02 Kr-85 1.635E+05 0.000E+00 2.067E+05 3.702E+05 Kr-85m 1.327E+04 0.000E+00 1.656E+04 2.983E+04 Kr-87 2.20 1EE+03 0.000E+00 2.780E+03 4.980E+03 Kr-88 1.592E+04 0.000E+00 2.009E+04 3.601E+04 Rb-86 4.741E-02 0.000E+00 6.660E-01 7.134E-01 Sr-89 6.063E-01 0.000E+00 2.139E+01 2.199E+01 Sr-90 5.960E-02 0.000E+00 2.127E+00 2.187E+00 Sr-91 9.835E-01 0.000E+00 1.973E+01 2.071E+01 Sr-92 4.235E-01 0.000E+00 4.621E+00 5.045E+00 Y-90 1.705E-03 0.OOOE+00 3.499E-01 3.516E-01 Y-91 8.108E-03 0.000E+00 3.294E-01 3.375E-01 Y-92 1.280E-01 0.000E+00 4.108E+00 4.236E+00 Y-93 7.808E-03 0.000E+00 1.602E-01 1.680E-01 Zr-95 1.094E-02 0.OOOE+00 3.867E-01 3.976E-01 Zr-97 2.965E-02 0.OOOE+00 7.148E-01 7.445E-01 Nb-95 1.099E-02 0.OOOE+00 3.921E-01 4.031E-01 Mo-99 1.398E-01 0.000E+00 4.309E+00 4.449E+00 Tc-99m 1.256E-01 0.000E+00 4.060E+00 4.186E+00 Ru-103 2.168E-01 0.000E+00 7.624E+00 7.841E+00 Ru-105 5.058E-02 0.000E+00 7.237E-01 7.742E-01 Ru-106 8.295E-02 0.000E+00 2.956E+00 3.039E+00 Rh- 105 7.076E-02 0.OOOE+00 2.102E+00 2.172E+00 Sb-127 1.548E-01 0.OOOE+00 4.934E+00 5.089E+00 Sb-129 3.147E-01 0.OOOE+00 4.439E+00 4.753E+00 Te-127 1.554E-01 0.OOOE+00 5.144E+00 5.299E+00 Te-127m 2.094E-02 0.OOOE+00 7.471E-01 7.680E-01 Te-129 3.651E-01 0.000E+00 6.53 1E+00 6.896E+00 Te-129m 7.001E-02 0.000E+00 2.464E+00 2.534E+00

I CALCULATION NO. H-1-ZZ-MDC-1880 IREVISION NO. 4 1 PAGE NO. 96 of 117 Table 18 (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 1.464E+00 0.OOOE+00 4.019E+0 1 4.166E+01 Te-132 2.098E+00 0.OOOE+00 6.585E+01 6.795E+01 1-131 2.907E+02 2.988E+04 3.574E+03 3.374E+04 1-132 3.351E+01 1.641E+02 2.924E+02 4.900E+02 1-133 1.096E+02 5.499E+03 1.1 82E+03 6.791E+03 1-134 3.023E+01 3.01 OE+0 1 3.735E+01 9.769E+01 1-135 6.693E+01 1.221E+03 5.329E+02 1.821E+03 Xe-133 6.786E+06 6.750E+04 8.538E+06 1.539E+07 Xe-135 1.377E+05 1.599E+04 1.748E+05 3.285E+05 Cs- 134 4.001E+00 0.OOOE+00 5.794E+0 1 6.194E+O I Cs-136 1.388E+00 0.OOOE+00 1.927E+01 2.065E+01 Cs-137 5.056E+00 0.OOOE+00 7.327E+01 7.833E+01 Ba-139 3.297E-0 I 0.OOOE+00 2.388E+00 2.718E+00 Ba-140 1.072E+00 0.OOOE+00 3.670E+01 3.777E+01 La-140 4.149E-02 0.OOOE+00 7.852E+00 7.893E+00 La-141 6.425E-03 0.OOOE+00 8.625E-02 9.267E-02 La-142 3.274E-03 0.OOOE+00 2.542E-02 2.869E-02 Ce- 141 2.559E-02 0.OOOE+00 8.979E-01 9.235E-01 Ce-143 2.248E-02 0.OOOE+00 6.280E-01 6.505E-01 Ce-144 4.187E-02 0.000E+00 1.491E+00 1.533E+00 Pr-143 9.258E-03 0.OOOE+00 3.338E-01 3.430E-01 Nd-147 4.056E-03 0.OOOE+00 1.380E-01 1.421E-01 Np-239 2.848E-01 0.OOOE+00 8.620E+00 8.905E+00 Pu-23 8 5.102E-05 0.OOOE+00 1.821E-03 1.872E-03 Pu-239 6.154E-06 0.OOOE+00 2.201E-04 2.262E-04 Pu-240 7.958E-06 0.OOOE+00 2.841E-04 2.920E-04 Pu-241 2.308E-03 0.OOOE+00 8.238E-02 8.469E-02 Am-241 1.039E-06 0.OOOE+00 3.743E-05 3.847E-05 Cm-242 2.459E-04 0.OOOE+00 8.746E-03 8.992E-03 Cm-244 1.183E-05 0.000E+00 4.222E-04 4.340E-04 A From HEPU300CLOO.oO B From HEPU300ES00.oO C From HIN300MSO0.oO

ICALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO.4 1PAGE NO.97 of 117 12.0 FIGURES Figure 1: Containment Leakage RADTRAD Nodalization

ICALCULATION NO. H-1-ZZ-MDC-1880 IREVISION NO.4 PAGE NO.98 of 117 I Figure 2: HCGS ESF Leakage RADTRAD Nodalization

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO. 4 PAGE NO. 9 of 117 A A>

         'A              A * >          A                           A
                                              /                    A Figure 3: HCGS MSIV Leakage Path Volumetric Distribution

ICALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO.4 1PAGE NO. 100 of 117

               'KKK       K~KKY~~* K K         '

I." K K' K K K V KKK

                                                                                              ~'
              *'~KK'                                                 K K     K>K ,K         K   .           -            K         - K      K
                                                                      'K'           -

Drywell V = 1.69E+05 ft3 Figure 4: HCGS MSIV Leakage RADTRAD Nodalization

ICALCULATION NO. H-1-ZZ-MDC-1880 IREVISION NO.4 1PAGE NO. 101 of 117 Figure 5 - HCGS Control Room RADTRAD Nodalization

CALCULATION NO. H-1-ZZ-MDC-1880 IREVISION NO. 4 PAGE NO. 102 of 117 EL 158'-9" CR Charcoal Filter Bed [Y] 3' (L) x 3' (H) x 4' (W) EL 155'-9" EL 155'-3" EL 154'-3"x CR Receptor Location (EL 143'-0") CT? fnrrrt~ncy VlPclnr (4 I l17'.*.n') Figure 6 - CR Filter Shine Dose (Elevation View)

ICALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO.4 PAGE NO. 103 of 117 X Indicates Dose Point Location Figure 7 - CR Filter Shine Dose (Plan View)

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO. 4 PAGE NO. 104 of 117 13.0 AFFECTED DOCUMENTS: HC Calculation H-1-ZZ-MDC-1880, Rev 3 will be superseded upon approval of this calculation. The following documents will be revised:

1. HCGS UFSAR Section 6.4

2. HCGS UFSAR Section 15.6.5.5.1

3. HCGS UFSAR Section 15.6.5.5.2

4. HCGS UFSAR Table 6.4-2

6. HCGS UFSAR Table 6.4-4

7. UFSAR Table 15.6-12

8. UFSAR Table 15.6-16

9. UFSAR Figure 15.6-3

10. Calculation H-I-ZZ-MDC-1923

11. Calculation H-1-ZZ-MDC-1927

12. Procedure VHC.MD-ST.GK-0002 14.0 ATTACHMENTS 14.1 E-mail

Subject:

FRVS Vent Charcoal Filter Efficiencies 14.2 Evaluation for Co-60 Isotope Test Assembly 14.3 1 CD with the following electronic files: Calculation No: H-1-ZZ-MDC-1880, Rev 4. RADTRAD computer runs

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO. 4 PAGE NO. 105 of 117 Attachment 14.1 From: Cichello, John P. Sent: Friday, March 15, 2002 10:40 AM To: Patel, Gopal J. Cc: Barkley, Barry L.; Duffy, John F.

Subject:

RE: FRVS Vent Charcoal Filter Efficiencies

Gopal, The CAV Units at Salem are tested for efficiency per ASTM 3803-89 at 30 degs C/95% RH (40 FPM with a 2" bed).

These parameters are the same conditions that the FRVS Vent Units will be revised to. Both systems will not have heaters and live in similar environments. The following is the CAV historical data. CAV 1 Penetration results CAV 2 Penetration results 1/8/99 0.508% 1/5/99 0.448% 9/28/00 0.542% 10/4/00 0.678% 1/28/02 0.805% Cichello Original Message ----- From: Patel, Gopal J. Sent: Monday, March 11, 2002 3:13 PM To: Cichello, John P. Cc: Barkley, Barry L.; Duffy, John F.

Subject:

FRVS Vent Charcoal Filter Efficiencies John: Please provide me a reference for the FRVS vent elemental and organic charcoal filter efficiencies of 90% without humidity control. As you told me that currently the Salem 1 & 2 charcoal is tested with 100% humidity with the in-laboratory methyl iodide penetration < 2.5%, which provides equivalent iodine removal efficiency of 95% or greater. Please provide me an E-mail response, which will attached as a reference for the same. I thank you very much for your cooperation and appreciate your quick response. Gopal J. Patel 1282 4.2 (pages 106 - 116) contains GEH Proprietary Information - It has been extracted from this document

ICALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO.4 1PAGE NO. 117 of 117 Attachment 14.3 1 CD With Various Electronic Files}}