Text

Calculation H-1-ZZ-MDC-1880, Revision 3, Post-LOCA Eab, LPZ and CR Doses, Attachment 5 to LR-N10-0163
	ML101390316
Person / Time
Site:	Hope Creek
Issue date:	09/02/2009
From:	Gita Patel; Public Service Electric & Gas Co
To:	; Office of Nuclear Reactor Regulation
References
	LR-N10-0163 CC-AA-309-1001, Rev 3, H-1-ZZ-MDC-1880, Rev 3, LS-AA-104-1001, Rev 2
	Download: ML101390316 (110)
	v • d • e

Attachment 5 LR-N10-0163 H-1-ZZ-MDC-1880, Revision 3, "Post-LOCA EAB, LPZ and CR Doses"

CC-AA-309-1001 Revision 3 ATTACHMENT 1 Design Analysis Major Revision Cover Sheet Design Analysis (Major Revision) I Last Page No, '109 Analysis No.: 4 H4-ZZ-MDC-6i880 Revision: t 3

Title:

Post.LOCA EAB, LPZ, and CR Doses ECIECR No.: NUCP 80096650go nr->r olr£ Revision:' 0 Station(s):' Hope Creek Component(s): -

Unit No.:' N NA Discipline:' NIA Descrip. Code/Keyword:" AST Safe'y/QAClass'." SR System Code: WNIA Structure'." NIA CONTROLLED DOCUMENT REFERENCES' _ _ _"_'_

Document No.: -ron/To Document No.: From/To

'H-I-ZZ-MDC-1879, Rev I From GE-NE-T230075M-OO-02 From 12-0025, Rev 3 From H-1-ZZ-MDC-0354, Rev I From ".','

GU-0013, Rev. 4 From' For remaining References see Section 10.0 From-Isthils Design Analysis Safeguards information?" yes [0 No Z Ifyes, see SY-AA-101-106 Does this Design Analysis cordain Unverified Assumplions?. ' Yes [I No ID If yes, ATI/AR#

This Design Analysis SUPERCEDES:"

fRvsin(itafce ,ecitoo atas:"Reiin3icroae Rev 2 HI-1-ZZ-MDC-1880, herdooia in itsmato entirety.

=,e Description of Revision {list affected pages for partials)":." Revision 3 incorporates the radiological impact of keeping the containment isolation valves open for 120 seconds fbllowing a design basis LOCA, Preparer Gopal J. Patel (NUCORE) 091011Z 009 PrliNa NI¢ Si-m lviý :#*J -- Date Method of Review.'" Detailed Review 0 Alternate Cac,1 chdTesting Reviewer. IMantk rur car (RIUCOREý 05)0212009 Review Notes:. Independent review fl Pe r review I]

IFcr Ex'I Afi*..

External Approven NA NI PSE&G Reviewers1" -John Duffyf~ijay Chandra ____________

independent Yd Party Review Required? 21 Yes 0 No.

PsE&G Approver:" James Boyar 1;Nl

.0Yk "-ýA r-'C , , V . LeuA W;'

i

A-TTACHMIENT 2 Owners Acceptance Review Checklist for External Design Analysis Page 1 of 1 DESIGN ANALYSIS NO. H-1-ZZ-MDC-1880 REV:

Yes No N/A

1. Do assumptions have sufficient rationale? 0 11 Are assumptions compatible with the way the plant is

2. operated and with the licensing basis? El El El .0

3. Do the design inputs have sufficient rationale? 0 El 4.. Are design inputs correct and reasonable? El Are design inputs compatible with the way the plant is operated and with the licensing basis?

El 11 9l

6. (Are Engineering Judgments clearly documented and justified?

Are Engineering Judgments compatible with the way the plant is operated and with the licensing basis?

El 0 19 Do the results and conclusions satisfy the purpose and objective of the Design Analysis?

Are the results and conclusions compatible with the way.

ol El 0 the plant is operated and with the licensing basis?

Does the Design Analysis include the applicable design basis documentation? 0g E3 El Have any limitations on the use of the results been

11. identified and transmitted to the appropriate 13 13 0 organizations?

12. Are there any unverified assumptions? El 0

13. Do allunverified assumptions have a tracking and closure El 1K mechanism in place?

Have a'11 affected design analyses been documented on the 14, Affected Documents List (ADL) for the associated 0l El 0 Configuration Change?.

Do the sources of inputs and analysis methodology used meet current technical requirements and regulatory commitments? (If the input sources or analysis El El El methodology are based on an out-of-date methodology or code, additional reconciliation may be required if the site has since committed to a more recent code)

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

17. Has the Vendor supplied the native electronic file(s)? [3l 0] 71 PSEG REVIEWER: J. Duffy/M.

Print / Sign Pentimall DATE: An*fW 11I6 y1ý D I CC-AA-309-1001, Rev 3

Comments:

1. Revision History: The summary for Revision 3 identifies that containment isolation closure is assumed at 120 seconds and ESF leakage is increased to 2.85 gpm. Briefly describe the changes in the analysis that offset the adverse impacts of the two changes identified.

> Incorporated. Revision History and Section 1.0 revised to acknowledge reduction in MSIV leakage and containment leakage doses by implementing models consistent with NRC approved models documented in recent BWR AST license amendments.

2. Include revision bars throughout the calculation to facilitate OWNER reviews.

> As stated in the Revision History, "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."

3. Section 2.1.2 indicates that two containment volumes are modeled; however; Figures 1, 4, and 9 all indicate one value of 169,000 cubic feet.

> A volume of 169,000 cf is input to the RADTRAD code. As noted in Sections 2.3.2 and 7.2, to account for the assumed mixing between the wetwell and drywell after 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and the resulting activity dilution, the flow rate through the MSIVs is reduced by the ratio of the drywell volume to thetotal volume at two hours. Section 7.11 also uses the larger volume in computing isotope removal after 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

4. Section 2.1.2 refers to amendments for other licenses, but no references are provided.

> Incorporated. Added AST license amendments for Dresden 2 & 3 (Ref. .10.70), Quad Cities 1 & 2 (Ref. 10.70), Peach Bottom 2 & 3 (Ref. 10.71), and Vermont Yankee (Ref. 10.72).

5. Section 2.7: The Section refers to Table 25, which list PCIVs. The valves are not individually modeled in the analysis. The table provides excessive detail that is not required for the analysis. Delete the table.

> The information in Table 25 is required to document that all PCIVs have been considered.

6. Section 2.7.2: Spell out abbreviation CPCS.

> Incorporated. Text has been added to define CPCS as the Containment Prepurge Cleanup System

7. Section 2.7.2: The section refers to the release through the PCIVs as bypass to the environment.

Appendix A of Regulatory Guide 1.183 includes the following guidance:

Primary containment leakage that bypasses the secondary containment should be evaluated at the bypass leak rate incorporated in the technical specifications. If the bypass leakage is through water, e.g., via a filled piping run that is maintained full, credit for retention of iodine and aerosols may be considered on a case-by-case basis. Similarly, deposition of aerosol radioactivity in gas-filled lines may be considered on a case-by-case basis.

Address whether there are any other bypasses.

> The purpose of this section is focused solely on the PCIVs, which provide a direct release path to the atmosphere bypassing the reactor building. The MSIV leakage path is the only licensing basis bypass leakage addressed in the analysis. If you feel that there are additional potential bypass leakage exist through other penetrations, please quantify them, document them in a design F Fr-7-_A A-*0r-1001. Rev 3

calculation, and inform the management to include in this analysis. This will be outside the scope of work of this revision.

8. Address the inconsistency between Design Inputs 5.3.2.11(9900 cfm) and 12 (9000cfm)

> There is no inconsistency. In Design Input 5.3.2.15 (last column) the FRVS exhaust rate given by the equation presented in Design Input 5.3.2.12 is multiplied by a factor of 1.1 to address the 10 percent uncertainty that is shown in Design Input 5.3.2.11. Design Input 5.3.2.12 is revised to acknowledge Design Input 5.3.2.15.

9. Provide a concise description of the modeling approach in the body of the analysis and as stated above summarize the changes from the previous model in the revised calculation in the revision history. For example, Section 7.10 and 7.11 concerns elemental iodine removal in containment, but there is no narrative explaining the modeling approach and why it is appropriate.

> Section 1.0 revised to acknowledge reduction in MSIV leakage and containment leakage doses by implementing models consistent with NRC approved models documented in recent BWR AST license amendments.

Section 2.1.3 refers to Section 7.11 when it describes the removal of elemental iodine by wall deposition on wetted surfaces inside containment, and that this removal process is modeled in accordance with Standard Review Plan 6.5.2. Section 2.1:3 is revised to include a discussion of the Section 7.10 discussion of the wetted drywell area that is used in the SRP 6.5.2 calculations, and the conservatism present in the area reduction percentages.

10. Section 7.2.1 indicates that MSIV leakage is increased from 46 scfh to 250 scfh, but the revision does not change MSIV leakage rate.

> Incorporated. The Section 7.2.1 text referring to 46 psig (sic) has been deleted.

1.1. Section 7.10 refersto an estimated 25%, 50%, and 75% of various areas, but provides no justification as to fractions being reasonable assumptions.

> Incorporated. Section 2.1.3 has been revised to include a discussion of the Section 7.10 discussion of the wetted drywell area that is used in the SRP 6.5.2 calculations, and the conservatism present in the area reduction percentages.

12. The statement in the conclusion that Therefore, pursuant to 10 CFR 50.59 guidance as defined in References 10.36 and 10.44, the proposed increase in the ESF leakage rate to 2.85 gpm, and the proposed increase in the maximum PCIV isolation time up to 120 seconds, and the resulting post-LOCA doses can be adopted as the current design and licensing bases for the HCGS. This is a 10CFR 5-0.59 conclusion and not an analysis conclusion.

> Incorporated. The Section 9.0 text presenting the 10CFR50.59 conclusion, has been deleted.

rC'..AA-*IOQ-101 'Rp~vl

ICALCULATIONNO. -- ZZMDC-1880 REVISIONNO.3 PAGE NO.5 of 109 REVISION HISTORY Revision Revision Description OIRO Initial Issue.

Revised due to incorporation of the preliminary plant-specific core OIR1 inventory, which will be confirmed via Order No. 80028003. The CR inleakage value was reduced to 900 cfm from 1000 cfm to offset the impact of preliminary core inventory on the CR dose.

Revised the aerosol removal rate in main steam piping, horizontal projected 0IR2 pipe surface area, and equation calculating the aerosol deposition.

Incorporated the revised x/Qs.

0 All interim revisions are incorporated. Isotopic activity released to environment is added. This is an 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 1 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.

CC-AA-309-1001, Rev 3 1

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO. 3 PAGE NO. 6 of 109 PAGE REVISION INDEX PAGE REV PAGE REV PAGE REV 1 3 42 3 84 3 2 3 43 3 85 3 3 3 44 3 86 3 4 3 45 3 87 3 5 3 46 3 88 3 6 3 47 3 89 3 7 3 48 3 90 3 8 3 49 3 91 3 9 3 50 3 92 3 10 3 51 3 93 3 11 3 52 3 94 3 12 3 53 3 95 3 13 3 54 3 96 3 14 3 55' 3 97 3 15 3 56 3 98 3 16 3 57 3 99 3 17 3 58 3 100 3 18 3 59 3 101 3 19 3 60 3 102 3 20 3 61 3 103 3 21 3 62 3 104 3 22 3 63 3 105 3 23 3 64 3 106 3 24 3 65 3 107 3 25 3 66 3 108 3 26 3 67 3 109 3 27 3 68 3 28 3 70 3 29 3 71 3 30 3 72 3 31 3 73 3 32 3 74 3 33 3 75 3 34 3 76 3 35- 3 77 3 36 3 78 3 37 3 79 3 38 3 80 3 39 3, 81 3 40 3 82 3 41 3 83 3 I CC-AA-309-1001, Rev 3

CALCULATION NO. H--ZZ-MDC-1880 TR EVISION NO.3 PAGE NO. 7 of 109 TABLE OF CONTENTS Design Analysis Cover Sheet Page Revision Index 6 Table of Contents 7 1.0 Purpose 8 2.0 Methodology 9 3.0 Acceptance Criteria 20 4.0 Assumptions 21 5.0 Design Inputs 25 6.0 Computer Codes & Compliance With Regulatory Requirements 33 7.0 Calculations 34 8.0 Results Summary 50 9.0 Conclusions 51 10.0 References 52 11.0 Tables 58 12.0 Figures 91 13.0 Affected Documents 101 14.0 Attachments 101 I CC-AA-309-1001,Rev 3

CALCULATION NO. H-I-ZZ-MDC-1880 - REVISION NO. 3 PAGENO. 8 of 109

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.

This calculation is revised to add an additional release path to evaluate the radiological impact of keeping the primary containment isolation valves (PCIVs) open for 120 seconds following a design basis LOCA, and to increase the allowable ESF leak rate from 1.0 gpm to 2.85 gpm. The 2.85 gpm ESF leak rate is modeled because the resulting total post-LOCA dose increase may be adopted as the Hope Creek licensing basis under 10 CFR 50.59 minimal dose increase criteria. 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.

4. Reactor Coolant System (RCS) Activity Release Via Open PCIV This calculation also includes the following changes:

The post-LOCA containment, ESF, and MSIV leakage path releases are reanalyzed to include progeny from the decay of parent radionuclides.

The MSIV leakage model has been updated for compliance with recent NRC Staff interpretations of the AEB 98-03 guidance. The updated model is consistent with AST license amendments for Dresden 2 & 3 (Ref. 10.70, Section 3.1.1.2) and Quad Cities 1 & 2 (Ref. 10.70, Section 3.1.1.2), and Peach Bottom 2 & 3 (Ref. 10.71, Section 3.2.2.8). The MSIV leakage model includes the following conservatisms:

Each MSIV release path consists of two well-mixed volume nodes consistent with AEB 98-03. This is believed to eliminate the potential variation of settling aerosol velocities in multiple volume nodes resulting from the different pressure/temperature boundary conditions and remove the in-series configuration of the aerosol and elemental iodihe removal efficiencies in the multiple volume nodes, which is believed to underestimate the resulting dose.

- The aerosol & elemental iodine removal is not credited in any steam line 96 hrs after the onset of a LOCA

- Aerosol & elemental iodine removal is not credited in the MSIV failed line between the reactor pressure vessel (RPV) nozzle and inboard MSIV for the entire duration of a LOCA, and

- The total MSIV leakage is distributed among two worst-case steam lines instead of four lines for thepurposes of this analysis.

- The aerosol settling velocity in the main steam lines beyond the outboard MSIVs is reduced to 30 percentile to justify that after having removed the heavier aerosol particles in the piping upstream of the outboard MSIV, the lighter aerosol particles do experience a reduced removal by gravitational deposition. The fine aerosol particles represented by a lower settling velocity, lesser deposition, and higher dose as well as coarser aerosol particles represented by a higher settling velocity, larger deposition, and lower dose.

CC-AA-309-1001, Rev 3

I CALCULATION NO. H-I-ZZ-MDC-1880 I REVISION NO. 3 I PAGE NO. 9 of 109 The containment leakage model has been updated to reduce the containment leakage activity is reduced by crediting 50% (mixing) dilution in the RB and removal by the Filtration, Recirculation, and Ventilation System (FRVS) filtration. The updated model is consistent with AST license amendments for Dresden 2 & 3, Quad Cities I & 2, Peach Bottom 2 & 3, and Vermont Yankee.

2.0 METHODOLOGY The design basis loss of coolant accident is analyzed using a conservative set of assumptions and as-built design inputs to demonstrate the performance of one or more aspects of the facility design to protect the control room operator and the health and safety of the general public. The guidance in Regulatory Guide 1.183 (Ref. 10.1) is followed along with the plant-specific design input parameters computable for the AST and TEDE dose criteria. The numeric values of the post-accident performance of ESF components are conservatively selected to assure an appropriate and prudent safety margin against unpredicted events in the course of an accident and compensate for large uncertainties in facility parameters, accident progression, radioactive material transport, and atmospheric dispersion.

2.1 Post-LOCA Containment Leakage:

2.1.1 Source Term:

The post-LOCA containment leakage model is shown in Figure 1. The average core inventory listed in Table ID is released into the containment at the release timing and fractions shown in RG 1.183 Tables 3 & 4 (Ref. 10.1, RGPs 3.2 & 3.3). Since the post-LOCA minimum suppression chamber water pH is greater than 7.0 (Ref. 10.43), the chemical form of radioiodine released into the containment is assumed to be 95% cesium iodide (CsI), 4.85 percent elemental iodine, and 0.15 percent organic iodide as shown in RG 1.183 RGP 3.5. With the exception of elemental and organic iodine and noble gases, the remaining fission products are assumed to be in particulate form (Ref 10.1., RGP 3.5). The plant-specific higher power core is developed in Table IA from Reference 10.45. As recommended in RADTRAD Table 1.4.3.3-2, the inventories listed for some of the parent isotopes include their significant daughter products. The final composite core inventory is shown in Table ID. The RADTRAD Nuclide Inventory File (NIF) HEPULOCA def.txt is developed based on the higher power core inventory in Table ID and used for the continment RCS Activity Release via open PCIV, ESF, and MSIV leakage paths. The source term design inputs are shown in Sections 5.3.1.1 through 5.3.1.8.

2.1.2 Transport In Primary Containment:

The radioactivity released from the fuel is assumed to mix instantaneously and homogeneously throughout the free air volume of the primary containment (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. The reduction in containment leakage activity by 50% (mixing) dilution in the RB and removal by the FRVS filtration is credited.

The analysis dilutes the radioactivity released from the core into the drywell air volume during the first 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of the LOCA, and then into the combined drywell plus suppression chamber air volume after 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , at which time the containment volume is expected to become well mixed following the restoration of core cooling because the thermal-hydraulic conditions in the primary containment are expected to be quite active due to a very high flow established between drywell and wetwell as a result CC-AA-309-1001, Rev 3

CALCULATION NO. HII-ZZ-MDC-1880 [REVISION NO.3 I PAGE NO. 10 of 109 of steaming and condensing phenomenon (Ref. 10.22, Table 2). The mixing over the remaining course of the accident (after 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ) has been accepted by the Staff in the previous AST license amendments for Dresden 2 & 3 (Ref. 10.70, Section 3.1.1.1), Quad Cities I & 2 (Ref. 10.70, Section 3.1.1.1), Peach Bottom 2 & 3 (Ref. 10.71, Section 3.2.2.2), and Vermont Yankee (Ref. 10.72, Section 3.2.1.2).

2.1.3 Reduction In Airborne Activity Inside Containment 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. Iodine removal by suppression pool scrubbing is not credited because the bulk core activity is released to containment well after the initial mass and energy release. Although containment sprays are not credited, the removal of the elemental iodine by wall deposition on wetted surfaces inside containment is modeled in accordance with SRP 6.5.2. 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 DF allows the elemental iodine to remain airborne in the drywell atmosphere for release to the atmosphere via containment and MSIV leakage. 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 surface area. The resultant modeled surface area of 33,200 ft is less than half of the available 69,126 ft2 drywell wetted surface area.

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).

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 3.7525E+20 atoms (Table 15).

4. The containment leakage case is analyzed in RADTRAD Run CUTOFF.oO using an elemental iodine removal rate of 3.16 hr1 for the first 2.0 hours0 days 0 hours 0 weeks 0 months of the accident, and then an elemental iodine removal rate of 1.74 hr-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.oO indicates the drywell elemental iodine atoms reach a value of 3.740E+20 by 4.0 hrs.

This means that an elemental iodine DF of 200 is reached in the drywell by 4.00 hrs. Elemental iodine wall removal is not credited in the analysis beyond this time.

2.1.4 Dual Containment:

Leakage from the primary containment is assumed to be released directly to the environment prior to draw down time during which the RB does not maintain a negative pressure as defined in technical CC-AA-309-1001, Rev 3

CALCULATION NO. H-,-ZZ-MDC-1880 REVISION NO.3 PAGE NO. 11 of 109 specifications (Ref 10.1, RGP A.4.2). 50% mixing is credited for dilution of the activity in the RB (Ref. 10.1, RGP A.4.4). The EAB, LPZ, and CR TEDE doses are shown in the Section 8.0.

2.1.5 Containment Purging:

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

However, the post-LOCA release through primary containment isolation valves (PCIVs) that are expected to remain open for 120 seconds following the onset of a design basis LOCA is analyzed in Section 2.7.

2.2 Post-LOCA ESF Leakage:

The post-LOCA ESF leakage release model is shown in Figure 2. The ESF systems that recirculate suppression pool water outside of the primary containment are assumed to leak during their intended operation. This release source includes leakage through valve packing glands, pump shaft seals, flanged connections, and other similar components. The radiological consequences from the postulated leakage is analyzed and combined with 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 RB.

2.2.1 Source Term:

With the exception of noble gases, all the fission products released from the fuel to the containment (as defined in Sections 5.3.1.3 & 5.3.1.5) are assumed to instantaneously and homogeneously mix in the suppression pool water at the time of release from the core. The total ESF leakage from all components in the ESF recirculation systems is assumed to be 2.85 gpm. This ESF leakage is doubled (Ref 10.1, Section A.5.2) and assumed to start at time t-=-0.0 minute after onset of a LOCA. With the exception of iodine, all remaining fission products in the recirculating liquid are assumed to be retained in the liquid phase. The design inputs for the ESF leakage are shown in Section 5.4.

2.2.2 Chemical Form The radioiodine that is postulated to be available for release to the environment is assumed to be 97%

. elemental and 3% organic (Ref. 10.1, RGP A.5.6). The reduction in ESF leakage activity by dilution in the RB and removal by FRVS recirculation and FRVS vent filtration systems are credited. The EAB, LPZ, and CR TEDE doses are shown in the Section 8.0.

2.3 Post-LOCA MSIV Leakage:

The MSIV leakage is postulated to release to the environment through the MSIV failed steam line and one of the three remaining intact steam lines. Each release path consists of two well-mixed volume nodes consistent with the AEB 98-03 (Ref. 10.22, Appendix A) two segment nodalization - piping between the inboard and outboard MSIVs and that between the outboard MSIV to Turbine Stop Valve (TSV) for the MSIV failed line and piping between the Reactor Pressure Vessel (RPV) nozzle to outboard MSIV and that between the outboard MSIV to TSV for Intact line. The well-mixed two volume nodes eliminate the potential variation of aerosol settling velocities in the multiple volume nodes resulting from the different temperature/pressure boundary conditions and remove the in-series configuration of the aerosol and elemental iodine removal efficiencies in the multiple volume nodes, which is believed to underestimate the resulting dose.

CC-AA-309-1001, Rev 3

CALCULATION NO. H-1-ZZ-MDC-1 880 _ REVISION NO.3 PAGE NO. 12 of 109 The post-LOCA MSIV Leakage model is shown in Figures 3 & 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. 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 flow carry any remaining fission products from the vessel, through the break, to the primary containment and provides new steam flow for rapid drywell-suppression air space mixing. The main steam isolation valves (MSIVs) are postulated to leak at a total design leak rate of 250 scfh at 50.6 psig. The radiological consequences from postulated MSIV leakage are analyzed and combined with the radiological consequences postulated for other fission product release paths to determine the total calculated radiological consequences from the LOCA (see Section 8.1 of this calc). The following assumptions are used for evaluating the consequences of MSIV leakage.

2.3.1 Source Term )

For the purpose of this analysis, the activity available for release via MSIV leakage is assumed to be that time dependent activity released into the drywell.

A total of 250 scfh MSIV leakage is assumed to occur as follows (see Section 2.3.2 for additional information regarding steam line selection):

(1) 150 scfh through the shortest steam line. This line is modeled as having the failed inboard MSIV.

Conservatively, 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. 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 0-24 hrs only.

(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 0-24 hrs only.

(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 lineis bounded.

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. 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 & 7.4.

The natural removal efficiency for elemental iodine in each steam line volume is assumed to be 50% as recommended in the AEB 98-03, Appendix B, page B-3. The post-LOCA time dependent MSIV leakage rates through the MSIV failed line and intact line are calculated in Section 7.2 and listed in Table 7.

CC-AA-309-1001, Rev 3

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO. 3 PAGE NO. 13 of 109 2.3.2 Determination of MSIV Leak Rates In Various Steam Line Volumes 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 & 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 aerosoi deposition. The main steam piping volumes and horizontal pipe surface are calculated in Section 7.3 and listed in Tables 2 through

4. The rate constant (%e) for different piping segments in the MSIV leakage release paths are calculated in Table 5 using 50 percentile aerosol settling velocity in piping segment between the RPV nozzle and outboard MSIV and 30 percentile aerosol settling velocity in piping segment between the outboard MSIV and TSV (Ref. 10.22, Appendix A, Table A-1) and applicable horizontal settling areas and volumes from Table 4. A lower aerosol settling velocity is used for the piping between the outboard MSIV and TSV to account for the lighter aerosol particle subject to a lesser removal due to gravitation.

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 MSIV leakage model and its technical basis is described below.

1. The MSIV failed line consists of two well-mixed nodes - the main steam line volume between the inboard MSIV & outboard MSIV as well as the main steam line volume between the outboard MSIV and turbine stop valves (TSV).

2. The intact steam line consists of two well-mixed nodes - the main steam line volume between the RPV and outboard MSIV as well as the main steam line volume between the outboard MSIV and TSV.

3. The 50th percentile settling velocity and resulting aerosol removal efficiencies are used in the first node of each line. This includes the MSIV failed line well-mixed volume representing the main steam line between the inboard and outboard MSIV, as well as the well-mixed volume representing the intact main steam line between the RPV and outboard MSIV.

4. The settling velocities in the main steam lines beyond the outboard MSIVs in both release paths (failed line and intact line) are reduced to 3 0 th percentile to account for a lesser deposition of lighter aerosol particles by gravitational deposition.

5. The trending of settling velocity with respect to percentile distribution is plotted in Figure 10 using the settling velocity information in Ref. 10.22, Appendix A, Table A-i, which indicates that the settling velocity is linearly proportional to percentile settling velocity distribution. This plot is used to extrapolate the intermediate 3 0 th percentile settling velocity.

The total MSIV leakage from all main steam lines is assumed to increase to 250 scfh measured at 50.6 psig, allowing a maximum of 150 scfh from any one of the 4 main steam lines. The total MSIV leak of 250 scfh is converted using the ideal gas law to determine the actual leakage (cfh) using the post-LOCA peak temperature and pressure in Section 7.2. Since the actual MSIV leak rate is reduced at the accident condition due to the combined effects of compression (due to the high pressure) and expansion (due to the high temperature), the increase in the MSIV leak rates to the environment from the TSVs are conservatively calculated in Section 7.2 using the Ideal Gas Law and drywell post-LOCA peak pressure and temperature and listed in Table 7. The MSIV leak rates in Table 7 are used in the analysis with CC-AA-309-1001, Rev 3

CALCULATION NO. H-I-ZZ-MDC-1880 REVISION NO. 3 I PAGE NO. 14 of 109 aerosol removal efficiencies calculated in Table 6 based on the horizontal pipe surface areas calculated in Section 7.3. To account for the assumed mixing between the wetwell and drywell after 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 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).

2.3.3 Recirculation Line Rupture Vs Main Steam Line Rupture Appendix A, "General Design Criteria for Nuclear Power Plants," to 10 CFR Part 50 defines LOCAs as those postulated accidents that result from a loss of coolant inventory at rates that exceed the capability of the reactor coolant makeup system. Leaks up to a double-ended rupture of the largest pipe of the reactor coolant system are included. The LOCA, as with all design basis accidents (DBAs), is a conservative surrogate accident that is intended to challenge selective aspects of the facility design. With regard to radiological consequences, a large-break LOCA is assumed as the design basis case for evaluating the performance of release mitigation systems and the containment and for evaluating the proposed siting of a facility. Therefore, a recirculation line rupture is considered as the initiating event rather than a main steam line rupture.

UFSAR Section 6.2.1.1.3.1 relating to the Pressure Suppression Containment System identifies that the containment functional design evaluation is based on the consideration of several postulated accidents, which include guillotine rupture of a recirculation line or main steam line. However, the UFSAR section indicates that maximum drywell and suppression chamber pressures occur as a result of the recirculation line break. Maximum drywell and suppression chamber pressure would result in maximum containment and MSIV leakage rates. Furthermore, UFSAR Section 6.2.1.1.3.3 states that following a main steam line break there is a rapid rise in water level. Therefore, a recirculation line rupture is considered to be the worst-case event with respect to radiological consequences. However, the steam line between the RPV nozzle and inboard MSIV in the MSIV failed line is postulated to be. ruptured to maximize the resulting dose by ignoringthe aerosol and elemental iodine removals.

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

The aerosol removal from the MSIV leakage is calculated in Section 7.4 using the NRC approved method in Reference 10.22, which uses the Monte Carlo distribution of aerosol settling velocity in well mixed flow. The analysis in Section 7.4 takes credit for the volumetric flow -rates and volumes of main steam piping upstream and downstream of outboard MSIVs because the steam lines from the RPV nozzle to turbine stop valve are seismically designed and supported for Safe Shutdown Earthquake (SSE) (Ref 10.26 & 10.37). The analysis in Section 7.4.1 determines that a large amount of airborne aerosol in MSIV leakage will be deposited on the steam pipe surface.

2.3.5 Elemental Iodine Removal The natural removal efficiency for elemental iodine in each steam line volume is assumed to be 50% as recommended in the AEB 98-03, Appendix B, page B-3.

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

Post-LOCA Containment Leakage CC-AA-309-1001, Rev 3

CALCULATION NO. H-1-ZZ-MDC-1880 [ REVISION NO. 3 I PAGE NO. 15 of 109

RCS Activity Release via open PCIV

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

Post-LOCA CREF filter shine 2.4.1 Post-LOCA Airborne Activity Inside CR The post-LOCA radioactive releases from various sources are discussed in Sections 6.1 through 6.3 above and shown in Figure 4. The 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 & 5.6.10 for the containment/ESF and MSIV leakages. The containment and ESF leakages have the same release point (FRVS vent) and X/Qs. The RADTRAD release models are developed for each release path using appropriate design inputs from Sections 5.3, 5.4, and 5.5. The CR dose model is developed using the design input parameters in Section 5.6. The CR airborne TEDE dose contributions from the above post-LOCA sources are calculated and tabulated in Section 8.0.

2.4.2 Post-LOCA Airborne Cloud External to CR The radioactive plumes released from various post-LOCA sources are carried over the CR building, submerging the CR in the radioactive cloud. The CR operator is exposed to direct radiation from the radioactive cloud external to the CR structure. The review of control building concrete structure drawings (Ref. 10.27 through 10.31) indicate that the CR is surrounded by at least 2'-i0-1/2" (1' ceiling

@ EL 155'-3" and 1'-10-1/2" roof @ EL 172'0") concrete shielding with a minimum distance of 29 feet from the least shielding (. 72'-0" - (13 7'-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.

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

2.4.4 Post-LOCA CREF Filter Shine The two trains of CREF charcoal and HEPA filters are located above the CR operating floor at elevation 155'-3" (Refs. 10.28, 10.29, & 10.39). The CR operating floor is located at elevation 137'-0" CC-AA-309-1001, Rev 3

CALCULATION NO. H-I-ZZ-MDC-1880 REVISIONNO.3 PAGE NO.16 of 109 (Ref. 10.28c). The concrete floor at EL 155'-3" is I feet thick (Ref. 10.29). The filter assembly is placed on a 6" concrete pad (Ref. 10.39c, Section DD), which provides the total concrete shielding of 1 '-6" between the CR operator and 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, &

10.39). Using full size drawings (Refs. 10.30 & 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 5, 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:

2.4.4.1 Post-LOCA Iodine & 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/[IEPA 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 17 based on the information obtained from RADTRAD run HEPU350CL02.o0. 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 17 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 18 using the atom/activity relation in Table 17.

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 16 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 19.

2.4.4.2 Post-LOCA Iodine & 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 (I{EPU35OES02.oO, 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 16 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 19.

CC-AA-309-1001, Rev 3

CALCULATION NO. H-I-ZZ-MDC-1880 7REVISION NO.3 PAGE NO. 17 of 109 2.4.4.3 Post-LOCA Iodine & 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 17 to calculate the aerosol isotopic activities deposited on the CREF BEPA 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 24 using the atom/activity relation in Table 17. 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 16 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 7 and Table 19.

2.4.4.4 MicroShield Analysis of CR Charcoal/HEPA Filter Shine The total CREF charcoal/HEPA filter iodine and aerosol isotopic activities in Table 19 are input into the MicroShield (Ref. 10.9) Computer Run HCCRFLT2.MS5 with the source geometry, dimension, and detector location as shown in Figure 6 & 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 & 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.

2.5 CR & 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 & 10.6). However, there is no specific criteria to establish the HEPA filter efficiency, the GL 99-02 criterion is used to determine the HEPA filter efficiency. The in-place penetration testing acceptance requirements are given in Hope Creek Technical Specifications (Ref. 10.6). The filter efficiencies credited in this analysis are calculated in Section 7.7 based on the testing criteria in Reference 10.6 and GL 99-02 (Ref. 10.3).

2.6 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 Amendment 134 (Ref. 10.49) 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 dose 1.183 limit.

The current licensing basis analysis is documented in the calculation H-1-ZZ-MDC- 1880, Rev 1. The increases in the proposed EAB, LPZ, & 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 F _ CC-AA-309-1001, Rev 3

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO. 3 PAGE NO. 18 of 109 the allowable regulatory guide limits. The comparisons in Section 8.2 confirm that,the proposed increases in the EAB, LPZ, & 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.

2.7 Reactor Coolant System (RCS) Activity Release Via Open PCIV The primary containment isolation valves (PCIVs) are expected to remain open for 120 seconds following the onset of a design basis LOCA. The release through these open PCIVs bypasses the reactor building and is directly released to the environment. The resulting doses from this release path at various receptor locations is calculated and added to the corresponding dose contributions from other post-LOCA releases in Section 8.0. All containment isolation valves associated with containment penetrations are listed in Table 25. The word "containment" is used in this calculation as a synonym for the drywell.

2.7.1 Regulatory Basis Per RG 1.183 (Ref. 10.1, Section 1.3.2), for the selected timing characteristics of the AST, e.g., change in the closure timing of a containment isolation valve, re-analysis of radiological calculations may not be necessary if the modified elapsed time remains a fraction (e.g., 25%) of the time between accident initiation and the onset of the gap release phase, which is 2 minutes or 120 seconds (Ref. 10.1, Table 4).

This regulatory guidance means that the NRC Staff,allows the licensees to increase the PCIV maximum isolation time to 30 seconds (0.25 x 120 seconds = 30 seconds) without reanalyzing of the design basis LOCA. For longer time delays, evaluation of the radiological consequences and other impacts of the delay, such as blockage by debris in sump water, may be necessary.

2.7.2 Radiological Evaluation The primary containment isolation valves listed in the Hope Creek Generating Station Technical Requirements Manual (HC TRM) Table 3.6.3-1 (Ref. 10.60) have been relocated from the HC Technical Specification by the 1-C operating license amendment No. 171 (Ref. 10.59) and their maximum isolation times are maintained in the HC TRM. The PCIV maximum isolation times are proposed to increase to 120 seconds. The PCIVs having a maximum isolation time of less than 120 seconds are listed in Table 25. A review of these PCIVs was performed based on the applicable P&IDs (Ref. 10.61) to determine whether any of these valves establishes a direct release path to the environment that bypasses the reactor building. The review indicates that the drywell purge exhaust (Penetration # 23, Isolation Valves GS-V024, V025, & V026) and suppression chamber purge exhaust (Penetration # 219, Isolation Valves GS-V027 & V028) (Refs. 10.60 and 10.61.10 & 10.61.15) could establish a direct release path to the environment during a LOCA. These purge exhaust isolation valves are proposed to remain open longer than 30 seconds (iLe., the isolation time which represents 25% of the time between the accident initiation and the onset of the gap release phase at 2 minutes); therefore, evaluation of the radiological consequences became necessary for 120 seconds closure time.

The Containment Prepurge Cleanup System (CPCS) fan provides a total of 3,000 cfm flow to the drywell and suppression chamber purge supply (Refs. 10.61.10 & 10.61.16). The maximum purge exhaust flow rate of 9,000 cfm is conservatively used in the analysis (Ref. 10.61.16). If the purge exhaust valves are kept open for 120 seconds, it establishes a release path to the environment either via the reactor building ventilation system exhaust tthough the south plant vent (Refs. 10.61.10 & 10.61.16)

CC-AA-309-1001, Rev 3

CALCULATION NO. H-1-ZZ-MDC-1880 I REVISION NO. 3 1 PAGE NO. 19 of 109 or through a 12" vent line (Ref. 10.6 1.10). The purge exhaust flow could also be released through the FRVS vent if purge exhaust is initiated by the LOCA signal.

As shown Figure 8, the vertical South Plant Vent pipe is located south-east of the CR air intake behind the reactor building (Ref. 10.13.h) with the release via the South Plant Vent line tip at EL 250'-0" (Ref. 10.13.h). The vent line tip is oriented such that the high energy release is further directed to the south of the CR air intake. Therefore, the use of FRVS vent x/Qs rather than the South Plant Vent X/Qs would yield less air dispersion and conservatively greater control room doses. The comparison of the FRVS vent and South Plant Vent is shown as follows:

FRVS Vent Release SPV Release Time CR* CR.

Interval X/Q X/Q (hr) (stm ') (s/m3)

A B 0-2 1.25E-03 5.75E-04 2-8 8.09E-04 3.84E-04 8-24 3.04E-04 1.40E-04 24-96 2.10E-04 9.08E-05 96-720 1.59E-04 7.01E-05 A From Ref. 10.5, Section 8.1 B From Ref 10.5, Section 8.4 The containment purge exhaust release is postulated to occur during a LOCA while the PCIVs remain open for 120 seconds. As shown in Figure 9, 100% of the radionuclide inventory inthe reactor coolant system (RCS) liquid is assumed to be released into the containment at the initiation of the LOCA (Ref. 10.1, Appendix A, Section 3.8). The RCS inventory is assumed to be instantaneously released and homogeneously mixed in the containment volume, which is consistent with the containment, leakage release assumption. The containment air is then released to the environment at a flow rate of 9,000 cfm for 120 seconds. The 120 second PCIV closure time is coincident with the onset of the gap release phase at 2 minutes (Ref. 10.1, Table 4). Therefore, the release fractions associated with the gap release and early in-vessel phases need not be considered as applicable during the purge exhaust release (Ref. 10.1, Appendix A, Section 3.8). The gap and early-in-vessel releases are considered in the containment leakage path (Section 2. 1).

The RCS activity is based on the technical specification for RCS specific activity of 0.2 ýiCi/g Dose Equivalent of 1-131 (Ref. 10.6.11). Iodine spikes are not considered (Ref. 10.1, Appendix A, Section 3.8). Per Technical Specification (TS) LCO 3.4.5 (Ref. 10.6.11), the specific activity of the primary coolant shall be limited to:

a. Less than or equal to 0.2 microcuries per gram DOSE EQUIVALENT 1-13 1, and

b. Less than or equal to 100/E microcuries per gram.

Technical Specification Section 1.11 (Ref. 10.6.20) defines DE 1-131 as that concentration of 1-131 (in

[tCi/g) which alone would produce the same thyroid dose as the quantity and isotopic mixture of 1-131, 1-132, 1-133, 1-134, and 1-135 actually present using the thyroid dose conversion factors (DCFs) listed in Table III of TID-14844 (Ref. 10.66).

CC-AA-309-1001, Rev 3

CALCULATION NO, H-1-ZZ-MDC-1880 . I PAGE NO. 20 of 109 I REVISION NO. 3 The plant-specific RCS iodine concentrations listed in Table 10 are obtained from Reference 10.69, Appendix A, and used with the iodine DCFs listed in Table 10 to establish an iodine scaling factor based on the iodine concentration of 0.2 .iCi/g DE 1-131. The iodine scaling factor is then used in Table I1Ito convert the normal iodine concentrations to 0.2 ýLCi/g DE 1-131. The isotopic noble gas concentrations are calculated in Table 12 using the noble gas release rate at time t = 0 sec (Ref. 10.68, Table V) and the steam mass flow rate of 17,774,000 lb/hr (Ref. 10.69, Section 3.2.1) in Section 7.12. The scaling factor needed to obtain I 00/S and resultant isotopic noble gas concentrations based on this 100/E-BAR scaling factor are calculated in Tables 13 and 14, respectively, using the following 100/E-BAR definition (Ref. 10.6.21):

E-BAR shall be the average, weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling, of the sum of the average beta and gamma energies per disintegration, in MeV, for isotopes, with half lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant (Ref. 10.6.21)..

The specific isotopic iodine and noble gas RCS activity concentrations are calculated in Table 15 using the RCS mass determined in Section 7.13, which are used to develop the RADTRAD NIF RCSdef.txt.

The RADTRAD dose conversion file (DCF) Fgrl 1&12.inp is modified to include short lived noble gas isotopes Kr-83m, Xe-131m, Xe-133m, Xe-135m, Xe-138 in the RCS inventory (See Table 12). The newly developed RCS def.txt and RCSFGR11 &12.txt files are used to calculate the CR and site boundary doses due to the containment purge exhaust release. The applicable additional assumptions and design inputs are shown in Sections 4 and 5. The RADTRAD input psf file for the containment purge exhaust release is HEPURCSCLOO.psf. The resulting doses are summed with the doses from other post-LOCA sources in Section 8.1.

The RCS iodine release into containment is modeled as 0% particulate, 97% elemental, and 3% organic.

This release profile has no impact on the resulting doses, because file HEPURCSCLOO.psf models the same iodine species activity removal efficiencies for all iodine removal mechanisms (i.e., control room charcoal and HEPA filters remove 99% of iodine; no spray or deposition iodine.removal mechanism is modeled).

Since the increase in dose is minimal, no changes are necessary to the containment shine, external cloud and control room filter shine doses calculated in Section 7.0.

3.0 ACCEPTANCE CRJTERIA 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. 10.67)

Dose Acceptance Criteria are:

Regulatory Dose Limits Dose Type Control Room (rem) EAB and LPZ (rem)

TEDE Dose 5 25 I CC-AA-309-1001, Rev 3

CALCULATION NO. H-1-ZZ-MDC-180 I N NO.3 PAGE NO.21 of 109

4.0 ASSUMPTIONS

The followhig 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 MWt (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 1 & 4).

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

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

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

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

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

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

4.6.3 The primary containment and the MSIVs are assumed to leak at the allowable Technical Specification peak pressure leak rate for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (Ref. 10.1, RGP A.3.7). This leakage is reduced to 50% of its TS value after the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months through day 30 (per Ref. 10.1, Sections 3.7 and 6.2). The flow is cut in half during these 29 days because the driving pressure after the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months averages less than 12.0 psig (that is, less than one-fourth of the Technical Specification peak pressure of 50.6 psig) and because flow is proportional to the square root of the pressure. The implied driving pressure for the 29-day period is 50.6 psig / 4 or 12.65 psig CC-AA-309-1001, Rev 3

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO.3 PAGE NO.22 of 109.

(27.35 psia). The post-LOCA pressure versus time curve for Case C in Reference 10.15 indicates that the pressure reaches a second peak of 15.8 psig at 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months and then decreases to the end of the curve at 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . UFSAR Figure 6.2-40 indicates that the pressure is 12.7 psig at 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , is less than 12 psig at 33 hours3.819444e-4 days 0.00917 hours 5.456349e-5 weeks 1.25565e-5 months , and then drops off rapidly at 42 hours4.861111e-4 days 0.0117 hours 6.944444e-5 weeks 1.5981e-5 months . UFSAR.

Figure 6.2-39 indicates that the average for the 29 days is only about 6 psig. (Note: If the KP system (MSIV Sealing System) deletion is credited in UFSAR Figure 6.2-39, the average for the 29 days is estimated to be only about 4 psig.) Thus the calculation is sufficiently conservative with respect to the 50% leakage rates based on the containment pressure behavior after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

4.6.4 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 Reference 10.42, SafetyEvaluation 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 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 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 hvdrogen vent and pury-e 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, However, the primary containment isolation valves (PCIVs) are expected to remain open for 120 seconds during a LOCA, which will introduce an additional .release path. The radiological impact of this release path through the open PCIVs is evaluated in Appendix A of this calculation and resulting .doses at various receptor locations are shown Section 8.0.

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 & 10.8).

CC-AA-309-1001, Rev 3

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO. 3 PAGE NO. 23 of 109 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 & 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 & Ref. 10.4).

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

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

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

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

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

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

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

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

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

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

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

4.8.2 The radioactivity releases and radiation levels used for the control room dose are determined using the same source term, transport, and release assumptions used for determining the CC-AA-309-1001, Rev 3

CALCULATION NO. H-1-ZZ-MDC-1880 [REVISION NO. 3 PAGE NO. 24 of 109 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 & 5.6.13 (Ref. 10.1, RGP 4.2.6).

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

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

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

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

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

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

Such a model would result in total unfiltered inleakage of 6,600 ft3 (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 ft3 (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).

4.8.8 The purge release evaluation should assume that 100% of the radionuclide inventory in the reactor coolant system liquid is released to the containment at the initiation of the LOCA (Ref. 10.1, Appendix A, Section 3.8).

4.8.9 The RCS inventory is assumed to be based on the technical specification reactor coolant system equilibrium activity; iodine spikes need not be considered (Ref. 10.1, Appendix A, Section 3.8).

4.8.10 The purge system prior to containment isolation should be analyzed and the resulting doses summed with the postulated doses from other release paths (Ref. 10.1, Appendix A, Section 3.8).

4.8.11 If the purge system is not isolated before the onset of the gap release phase (i.e., 2 minutes per Ref. 10. 1, Table 4), then the.release fractions associated with the gap release and early in-vessel phases should be considered as applicable (Ref. 10.1, Appendix A, Section 3.8)

CC-AA-309-1001, Rev 3 ]

CALCULATION NO. H-l-ZZ-MDC-1880 REVISION NO. 3 PAGE NO. 25 of 109 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 groun.d 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 MSTV 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 y,/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 CC-AA-309-1001, Rev 3

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO. 3 PAGE NO. 26 of 109 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.

I CC-AA-309-1001, Rev 3

CALCULATION NO. H-1-ZZ-MDC-1880 -REVISION NO. 3 PAGE NO. 27 of 109 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 Post-LOCA Containment Condition (Ref. 10.15) 0-0.5 hr (Cont. Pressure) 63 psia 10.15 0.5-720 hr (Cont. Pressure) 31 psia 5.3.1.3 Isotopic Average Core Inventory (Ci/MWt) (Table ID) (Ref. 10.45) See Note Below Isotope Ci/MW, Isotope Ci/MWt Isotope Ci/MWI 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 3.300E+02 RU106 2.940E+04 BA139 4.950E+04 KR 85M 7.350E+03 RHI05 2.530E+04 BA140 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 TE127M 2.780E+03 LA142 4.370E+04 SR 89 2.690E+04 TE127 3.710E+02 CE141 4.540E+04 SR 90 2.640E+03 TE129M 8.350E+03 CE143 4.220E+04 SR 91 5.300E+04 TE129 1.240E+03 CE144 7.424E+04

SR92 3.610E+04 TE131M 2.764E+04 PR143 4.080E+04 Y90 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 Y93 4.160E+04 1133 5.510E+04 PU23 9 1.090E+01 ZR95 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 CS134 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 1C 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 Early In-Vessel Release 0.5 hr 1.5 hr 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%

CC-AA-309-1001, Rev 3

CALCULATION NO. H-1-ZZ-MDC-1880 TREVISION NO. 3 PA'GE NO. 28 of 109 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 Bumup 58 GWD/MTU < 62 GWD/MTU 10.54 & 10.1 5.3.2 Activity Transport in Primary Containment 5.3.2.1 Primary Containment Parameters 5.3.2.2 Drywell Air Volume 169,000 ft3 10.6.6 & 10.16 5.3.2.3 Suppression Chamber Air 137,000 ft3 10.6.6 & 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 0-24 hrs [0.5 v%/day 10.6.4 & 10.15 24-720 hrs 0.25 v%/day 10.1, RGP A.3.7 & 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 ft . 10.6.7 5.3.2.10 Reactor Building Mixing 50% 10.1, RGP A.4.4 5.3.2.11 FRVS Vent Exhaust 9000 cfm +/- 10% 10.6.3, & 10.61.16 Rate Before Draw Down 5.3.2.12 FRVS Vent Exhaust 3324 + 5676el1$t Actual Eqn in Ref. 10.19, page 24 Flow Rate After Draw Down Flow Rates calculated in Design is 3324 + 5 6 3 7 e118t Input 5.3.2.15 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 CC-AA-309-1001, Rev 3

CALCULATION NO. H--ZZ-MDC-1880 REVISION NO. 3 PAGE NO.29 of 109 Design Input Parameter Value Assigned Reference 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 + 5676e"1'18t 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 cfm) -10% 10.6.12 & 10.61.16 (or, 108,000 5.4 ESF Leakage Model Parameters 5.4.1 Sump Water Volume 1.18,000 ft3 10.6.5 & 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 & Tables 1 & 4)

Group Gap Release Phase Early In-Vessel Release Phase Timing Duration (Hrs) 2 min - 0.50 Hr 0.50 - 2.0 Hr Halogen 0.05 0.25 5.4.6 Iodine Flashing Factor 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 4-1 5.5 MSIV Leakage Model Parameters 5.5.1 Total MSIV Leak Rate <250 scfh 10.6.17 Through All Four Lines 5.5.2 MSIV Leak Rate Through 150 scfh 10;6.17 Line With MSIV Failed 5.5.3 MSIV Leak Rate Through 50 scfh Assumed First Intact Line 5.5.4 MSIV Leak Rate Through 50 seth Assumed Second Intact Line 5.5.5 Number of Steam Lines 4 10.11 & 10.12e 5.5.6 Diameter and Wall Diameter = 26" 10.13b Thickness of Pipe Between RPV Wall Thickness = 1.117" 10.14c Nozzle & Inboard Isolation Valves HV F022A/B/C/D CC-AA-309-1001, Rev 3

CALCULATION NO. H-1-ZZ-MDC-1880. REVISION NO. 3 J PAGE NO.30 of 109 Design Input Parameter Value Assigned Reference 5.5.7 Diameter and Wall Diameter = 26" 10.12e Thickness of Pipe Between Wall Thickness 1.117" 10.14c Inboard & Outboard Isolation Valves HV F028A/B/C/D 5.5.8 Diameter and Wall Diameter - 26" 10.12e Thickness of Pipe Between Wall Thickness 1.023 10.14a Outboard & 3rdIsolation Valves I-HV 363 1AIB/C/D 5.5.9 Diameter of Pipe Between Diameter.= 28" 10.12a 3rd Isolation & Turbine Stop. Wall Thickness = 0.934" 10.14b Valves MSV 1/2/3/4 5.5.10 Corrosion Allowance For 0.12" 10.14 Steam 5.5.11 Drywell Peak Pressure 50.6 psig 10.17, Attachment 1, PUSAR Table 4-1 5.5.12 Drywell Peak 298uF 10.17, Attachment 1, PUSAR Temperature Table 4-1 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 5.6.3 CR Minimum Recirculation 2,600 cfm 10.6.15 Flow Rate .

5.6.4 CR Unfiltered Inleakage 196 +/- 1.0 cfm actually measured 10.46, page 56 After CREV System Initiation 350 cfm Assumed 5.6.5 CREV System Initiation 30 minutes Assumption 4.8.5 Time After a LOCA 5.6.6 CR Charcoal & IEPA 99% Sections 7.7 & 7.9 Filter Efficiencies 5.6.7 CR Unfiltered Inleakage 500 cfm 10.40, page 6.4-8 & Assumption Prior to CREV System Initiation 4.8.6 5.6.8 CR Concrete Wall, Floor, and Ceiling Thickness .....

Walls >3 feet 10_27 through 10.31 Floor >3 feet Total Roof Thickness T-10-1/2" Ceiling Above CR I '-0" _ _.... 10.29a & 10.29b 5.6.9 CR X/Qs For Containment & ESF Leakage Release Via FRVS Vent Ground Level Release Time X/Q (see/rn) 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 ..... _ _

CC-AA-309-1001, Rev 3

CALCULATION NO. H-1-ZZ-MDC-1880 FREVISION NO. 3 PAGE NO.31 of 109 Design Input Parameter Value Assigned Reference 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 XIQs For MSIV Leakage Release Via Turbine Building Louvers Ground Level Release Time X/Q (sec/mr3) 0-2 6.17E-04 10.5, page 35 2-8 4.OOE-04 8-24 1.44E-04 24-96 1.OOE-04 96-720 7.49E-05 5.6.12 CR Breathing Rate 3.5E-04 (m3/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/m 3 10.32, pages 5 & 9 5.7.2 EAB Breathing Rate 3 3.5E-04 m /sec 10.1 5.7.3 LPZ X/Qs (0-720 Hrs)

Time X/Q (see/mr) 0-2 1.9E-05 10.32, pages 5 & 9 2-4 1.2E-05 4-8 8.OE-06 8-24 4.OE-06

.24-96 1.7E-06 96-720 4.7E-07 5.7.4 Offsite Breathing Rates Time BR (m3/see) 0-8 3.5E-04 10.1, RGPs 4.1.3 & 4.4 8-24 1.8E-04 24-720 2.3E-04 5.7.5 CR Charcoal Filter Dimensions Approximated Conservatively Length 3 feet 10.38 Height 3 feet Width 4 feet 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" 5.8 Containment Purge Exhaust Release Parameters 5.8.1 Maximum PCIV Closure 120.0 sec Assumed (Isolation) Time ._

5.8.2 Containment Purge Exhaust 3,000 cfm 10.61.16 Rate 9,000 cfm Used in analysis CC-AA-309-1001, Rev 3

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO. 3 PAGE NO. 32 of 109 Design Input Parameter Value Assigned Reference 5.8.3 Reactor Coolant System 13,099 ft3 Section 7.13 Volume.

5.8.4 Uprated Steam Flow Rate 17,774,000 lb/hr 10.69, Section 3.2.1 5.8.5 RCS Volume Of Liquid In 11,885 ft3 10.63 Vessel.

5.8.6 RCS Volume Of Liquid In 1,214 ft3 10.63 Recirculation Loops 5.8.7 RCS Volume Of Steam In. 9,089 ft3. 10.63 Vessel __

5.8.8 Total Water And Steam. 21,970 ft3 10.6.22 Volume Of Reactor Vessel &

RCS 5.8.9 Iodine Specific Activity 0.2 tCi/g DE 1-131 10.6.11 5.8.10 Noble Gas Specific I 00f/s Ci/g 10.6.11 Activity 5.8.11 Maximum RCS Noble Gas Release Rates (Ret. 10.68, Table V)& Iodine Activity (Ref. 10.69, Appendix A)

[Note: Ref. 10.68 Table V noble gas release rates at time equal zero (no decay) are conservatively greater than or eaual to the Ref. 10.69, Anpendix A, noble gas release rates..]

Isotope ýiCi/sec Isotope gCi/sec Isotope g.Ci/g KR-83M 3.400E+03 XE-133M 2.900E+02 1-131 1.300E-02 KR-85M 6.100E+03 XE-133 8.200E+03 1-132 1.200E-01 KR-85 2.OOOE+01 XE-135M 2.600E+04 1-133 .8.900E-02 KR-87 2.OOOE+04 XE-135 2.200E+04 1-134 2.400E-01 KR-88 2.00E+04 XE-138 8.900E+04 1-135 1.300E-01 XE-131M 1.500E+01 CC-AA- 09-1 001, Rev 3

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO. 3 PAGE NO. 33 of 109 6.0 COMPUTER CODES & 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:

a 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 PSE&G in various AST license amendments, which are approved by the NRC. PSE&G 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.

e MicroShield 5.05 (Ref. 10.9): A commercially available and accepted code used to deteirmine 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 PSE&G in various AST license amendments, which are approved by the NRC. PSE&G 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.

6.2 Compliance With Re-ulatorv Requirements As discussed in Section 4.0, Assumptions, the analysis in this calculation complies with line-by-line requirements in Regulatory Guide 1.183.

I CC-AA-309-1001, Rev 3

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO.3 PAGE NO. 34 of 109 7.0 CALCULATIONS 7.1 HCGS Plant Specific Nuclide Inventory File (NIF) For RADTRAD3.02 Input The RADTRAD nuclide inventory file Bwr def NIF establishes the power dependent radionuclide activity in Ci/MWt for the reactor core source term. Since these core radionuclide activities are dependent on the core thermal power level, reload design, and burnup, the NIF is modified based on the plant-specific core inventory information obtained from Reference 10.45. The RADTRAD NIF HEPULOCA_DEF.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 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , and reduced in half to 0.25 w%/day after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . The total containment leakage does not include the leakage through the MSIVs.

7.2.2 MSIV Leakage Durine 0-24 hrs Note: The RADTRAD runs model MSIV leakage beginning at 2 minutes, which is coincident with the start of the gap release per Section 5.3.1.5.

Total Drywell volume = 169,000 ft3 (Ref. 10.16)

Total MSIV leakage measured g 50.6 psig 250 scffi 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, Attachment 1, PUSAR Table 4-1)

T @LOCA = Drywell peak temperature = 2980 F (Ref. 10.17, Attachment 1, PUSAR Table 4-1) = 298'F

+ 460 = 758R P @STP = Standard pressure = 14.7 psia T @STP = Standard temperature = .68F = 68°F + 460 = 528°R V @STP = MSIV leakage based @ 50.6 psig =.250 scfh V @LOCA = (PV/T @STP) x (TiP @LOCA) 0-2 hrs MSIV leakage @ drywell peak pressure of 50.6 psig and temperature of 298OF

= 250 scfh x [14.7 psia / (50.6 psig + 14.7 psia)] x [758'R / 528°R]

= 250 scfh x 0.225 x 1.436 = 80.78 cfh

= (80.78 ft3/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 scff 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.

CC-AA-309-1001, Rev 3

CALCULATION NO. H-1-7Z-NMDC-1880 REVISION NO.3 PAGE NO. 35 of 109 0-2 hrs allowable leakage from the MS line with a failed MSIV (at maximum 150 scfh leak rate)

= (150 scfh / 250 scfh 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 2-24 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) 2-24 hrs MSIV leakage @ drywell peak pressure of 50.6 psig = 80.78 cfh (Section 7.2.2).

=(80.78cfh x 24 hr/day) x 100% /.3.06E+05 ft = 0.634 %/day Corresponding MSIV leak rate = 80.78 cfh x (i.69E+05 ft3 / 3,06E+05 ft3) = 44.61 cfh 2-24 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 2-24 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 MSIV Leakage Durina 24-720 hrs The total MSIV leakage is reduced by 0.50 due to the reduction in the drywell pressure (Ref. 10.17, Section .2.1.2.3)

Total MSIV Leakage = 0.50 x 250 scfh = 125 scflh MSIV leakage in failed steam line = 0.50 x 150 scfh 75.0 scfh MSIV leakage in intact steam line = 0.50 x 100 scfh = 50.0 scfh Corresponding MSIV leak rate = 80.78 / 2 (Section 7.2.3) = 40.39 cfh 24-720 hrs MSIV leakage (40.39 cfh x 24 hr/day) x 100% / 3.06E+05 ft3 = 0.31.7 %/day Corresponding MSIV leak rate 40.30 cff x (1.69E+05 ft3 / 3.06E+05 ft3) = 22.31 cfh 24-720 hrs allowable leakage from the MS Line with a failed MSIV (at maximum 75 scfh leak rate)

= (75 scfh / 125 scth total) x 22.31 cfh = 13.39 cfh = 0.223 cfm 24-720 hrs allowable leakage from the shortest intact MS Line (at maximum 50 scfh leak rate)

= (50 scfh 125 scfh total) x 22.31 cfh = 8.92 cfh = 0.149 cfm I CC-AA-309-1001,.Rev 3

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO.3 PAGE NO. 36 of 109 7.2.5 MSIV Leakage To Environment:

7.2.5.1 MSIV Leakage into the pipe spool between outboard MSIV & TSV and environment from MSIV failed line (MS Line 1) 0-24 hrs It is assumed that the post-LOCA activity released in the steam line (SL) with the failed inboard MSIV is instantaneously and homogeneously distributed in the single volume of SL between the RPV nozzle and outboard MSIV (Volume V1 = V11 + V12) (well mixed volume), where Volume V, 1 is the SL volume between the RPV nozzle and the failed inboard MSIV, and Volume V12 is the SL volume between the failed inboard MSIV and the outboard MSIV. Volume V13 defines the SL volume from the outboard MSIV to the Turbine Stop Valve (TSV).

It is conservatively assumed that the MSIV leakage past the outboard MSIV (i.e., from volume V1 between the RPV and the outboard MSIV to volume V13 between the outboard MSIV and the TSV) expands to the atmospheric condition as follows:

Upstrearm of outboard MSIV in MSIV failed line (Section 7.2.2):

VI 48.47 cfh .P1- 50.6 psig + 14.7 = 65.3 psia TI (2980 F + 460) 758'R Downstream of outboard MSIV in MSIV failed line (Atmospheric Condition):

V2 TBD P2 14.7 psia T2 = (68'F + 460) = 5280R MSIV Leakage into the pipe spool between outboard MSIV & 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'R) x (528°R / 14.7 psia) 150 cfh =2.5 cfm This is as expected, given that the 48.47 cfh leakage rate is equivalent to 1.50 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.24-720 hrs The total MSIV leakage is reduced by 0.50 due to the reduction in the drywell pressure (Ref. 10.17, Section 2.1.2.3)

[13.39 cfh (Section 7.2.4) / 26.77 scfh (Section 7..2.3)] x 150 cfh = 75.00 cfh = 1.25 cfm 7.2.5.2 MSIV Leakage into the pipe spool between outboard MSIV & TSV and environment from MSIV shortest intact line (MS Line 2) 0-24 hrs Upstream of inboard MSIV in the shortest intact MS Line (Section 7.2.2):

V1 = 32.31 cfh P 1 50.6 psig + 14.7 = 65.3 psia Ti = (298'F + 460)= 758'R CC-AA-309-1001, Rev 3

I CALCULATION NO. H-I-ZZ-MDC-1880 I REVISION NO. 3 PAGE NO. 37 of 109 CLUAINN.HI-ZDC88 REISINN. PAG NO 7oI0 Downstream of inboard MSIV in intact line (assumed Atmospheric Condition):

V2 = TBD P2 = 14.7 psia T2 = (689F + 460) = 528°R MSIV Leakage into the intact pipe spools between the inboard & 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'R) x (528°R /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 & 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.24-720 hrs The total MSIV leakage is reduced by 0.50 due to the reduction in the drywell pressure (Ref. 10.17, Section 2.1.2.3)

[8,92 cfh (Section 7.2.4) / 17.84 scfh (Section 7.2.3)] x 100 cfh 50.00 cfh = 0.833 cfm 7.3 Main Steam Line Volumes & Surface Area for Plateout of Activity The two shortest main stream lines are selected for the aerosol deposition, namely the steam lines "A"

& "D" connected to the reactor pressure vessel (RPV) nozzle N3A & 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, & 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 & 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)

Total Minimum Thickness = 1.117" + 0.12" = 1.237" 26" Pipe ID = OD - (2 x Min Wall Thickness) = 26" - 2 x 1.237" = 23.526" 1.961" Pipe Flow Area = (7c/4) x (Pipe I.D.) 2 = (3.14 / 4) x (1.961 ft)2 = 3.019 ft2 Length of short radius (SR) elbow = Re 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)

CC-AA-309-1001, Rev 3

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO. 3 PAGE NO. 38 of 109 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 & Third Isolation Valves Piping Class DBB (Ref. 10.11 & 10.12e)

Pipe Diameter= 26" (Ref. 10.12e)

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

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

Total Minimum Thickness = 1.023" + 0.12" = 1.143" 26" Pipe ID = OD - (2 x Min Wall Thickness) = 26" - 2 x 1.143" = 23.714" = 1,976' Pipe Flow Area = (n/4) x (Pipe I.D.) 2 = (3.14 / 4) x (1.976 ft)2 3.065 ft2 7.3.1.3 MSIV Line Between Third Isolation Valve and Turbine Stop Valve Piping Class= DBC (Ref. 10.11 & 10.12a)

Pipe Diameter = 28" (Ref. 10.12a)

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

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

Total Minimum Thickness =0.934" +.0,12" 1.054" 28" Pipe ID= OD - (2 x Min Wall Thickness) = 28" - 2 x 1.054" 25.892" = 2.158' ft2 Pipe Flow Area = (n/4) x (Pipe I.D.)2 = (3.14 / 4) x (2.158 ft)2 = 3.656 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 & 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):

Length of Pipe Between Inboard & Outboard Isolation Valves Distance between the RPV centerline and inner edge of inboard 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)

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'-l0" + 5'-3"= 9'-!"

Distance between inner edge of inboard MSIV and outer edge of outboard MSIV

= 24'-8" + 9'-1"= 33.75' Total Pipe Volume = Flow Area x Total Length V1 = 3.019 ft2 x 33.75 ft 101.89 ft3 Total horizontal pipe surface area = D x L (Horizontal Length)

CC-AA-309-1001, Rev 3

CALCULATION NO. H-1-ZZ-MDC-1.880 REVISION NO. 3 PAGE NO. 39 of 109

= 1.961 ftx 33.75 ft= 66.18 ft2 3 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 f 2 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' 3 2

Vertical volume V = Flow area x Length = 3.065 ft x 17.375 ft = 53.25 ft SR elbow volume = 10.27 ft 3 (Section 7.3.1.1)

Horizontal pipe = 6'-0" + 5'-3" = 11.25' 3 Horizontal volume V = Flow area x Length = 3.065 ft2 x 11.25 ft = 34.48 ft 3

Total Pipe Volume = 17.62 ft3 + 10.27 ft3 + 53.25 ft3 + 10.27 ft3 + 34.48 ft = 125.89 ft 3

Total horizontal pipe surface area = D x L (Horizontal Length)

= 1.976 ft x (5.75'+ 11.25') = 33.59 ft2 Total horizontal pipe volume,= 17.62 ft3 + 34.48 ft3 = 52.10 ft3 7.3.2.3 Piping from Third MSTV 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' 2

Horizontal volume Flow area x Length = 3.656 ft x 137.75 ft = 503.61 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 = 14'-1 1" +.38'-0" + 38'-0' - 28" = 88'-7" = 88.583' Horizontal volume = Flow area x Length = 3.656 ft 2 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' Horizontal volume Flow area x Length = 3.656 ft2 x 43.458 ft 158.88 f 3 Total pipe volume 503.61 ft3 + 13.40 ft3 + 323.86 ft3 + 13.40 ft3 + 158.88 ft 3 =1,03.!5 ft, Total horizontal pipesurface 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 ftx 277.123 ft= 598.03 ft Total horizontal pipe volume = 1,013.15 ft3 7.3.2.4 Volume & Surface Area for Aerosol Deposition - Outboard MSIVto TSV MSV 3 Combined total pipe volume V2 = Outboard MSIV to Third MSIV + Third 3MSIV to TSV MSV 3

= 125.89 ft3 (Section 7.3.2.2) + 1,013.15 ft3 (Section 7.3.2.3) = 1,139.04 ft Combined total horizontal pipe surface area

= 33.59 ft 2 (Section 7.3.2.2) + 598.03 ft 2 (Section 7.3.2.3) = 631.62 ft2 Total horizontal pipe volume

= 52.10 ft3 (Section 7.3.2.2) + 1,013.15 ft 3 (Section 7.3.2.3) = 1,065.25 ft3 7.3.3. Piping Volume & 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):

CC-AA-309-1001, Rev 3

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO. 3 PAGE NO. 40 of 109 Nozzle elevation (Center Line) 170'-1-1/2" 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' = 1.87' Volume = 3.019 ft x 1.87 ft = 5.66 ft3 LR Elbow Volume = 15.43 ft3 (Section 7.3.1.1)

SR elbow volume = 10.27 ft3 (Section 7.3.1.1)

Height of 10'-10" Bend = 1'-7/14" / Sin 750 ~ 12'-0" Elevation of bend =.123'-6" + 12'-0" = 135'-3" Vertical Pipe = 170'-14/2" - 135'- 3"- 2'2" = 32'- 8-1/2" 32.708' Vertical Volume, V = 3.0193 ft2 x 32.708 ft = 98.75 ft3 SR elbow V = 10.27 ft Length of bend =1 1'-7-1/4" - 5.11 = 6.494' Vertical Volume, V = 3.019 3 ft2 .x 6.494' = '19.61 ft'3 LR elbow V = 15.43 ft Horizontal bend pife = 4'-4-15/16 + 9'-2-3/8" = 13.609'

Volume = 3.019 ft x 13.609 ft = 41.09 ft3 3

LR elbow V 15.43 ft Vertical pipe 123'-5-1/2" - 107'-0-3/8" - 39" - 13.18' Vertical volume = 3.019 2 3 ft x 13.18 ft = 39.79 ft3 LR elbow V = 15.43 ft Horizontal length = 4'-7" - 39." = 1.33' Horizontal volume = 3.019 ft2 x 1.33 ft = 402 ft3 3 2

Horizontal LR elbow = 3.019 ft x 5.11 ft = 15.43 ft Horizontal length = 3'-6-1/4" = 3.521' Horizontal volume = 3.019 ft2 x 3.521 ft= 10.63 ft 3 Total Volume of Steam Header From RPV Nozzle A To Inboard MSIV AO-80A 2VATI

= 5.66. ft3 + 10.27 ft3 + 98.75 ft3 + 10.27 ft3 + 19.61 ft3 + 15.43 ft3 + 41.09 ft' + 15.43 ft3 + 39.79 ft3 +

3 4.02 ft3 + 15.43 ft3 + 10.63 ft3 = 301.81 ft3 15.43 ft'+

Total horizontal pipe length 1.87' + 13.609' + 1.33' + 5.11' + 3.521' = 225.44' Total horizontal pipe surface area = D x L = 1.961 ft x 25.44 ft = 49.89 ft Total horizontal pipe volume 5.66 ft3 + 41.09 ft3 + 4.02 ft3 + 15.43 ft3 + 10.63 ft3 = 76.83 ft3 7.3.3.2 Piping from Inboard MSIV V031 to Outboard MSIV V035 (Ref. 10.12.e & 10.21.b):

Same as in Section 7.3.2.1 3

Total Pipe Volume = Flow Area x Total Length = 101.89 ft Total horizontal pipe surface area = D x L3 (Horizontal Length) 66.18 ft2 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 Total Pipe Volume = 125.89 ft3 Total horizontal pipe surface area = D x L (Horizontal Length) 33.59 ft2 CC-AA-309-1001, Rev 3

PAGE NO. 41 of 109 -

CALCULATION NO. H-1-ZZ-MDC-1880 I REVISION NO. 3 Total horizontal pipe volume = 52.10 ft3 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" = 131.833' Horizontal volume = Flow area x Length = 3.656 ft2 x 131.833 ft = 481.98 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 = 6'7" + 40'-0" + 43'6" - 28" = 87'-9" 2

= 87.75' 3 Horizontal volume = Flow area 3x Length = 3.656 ft x 87.75 ft = 320.81 ft Horizontal SR elbow = 13.40 ft Horizontal pipe = 26'-4-1/2" + 23'-0" = 49'-4-1/2" = 49.375' Horizontal volume.= Flow area x Length = 3.656 ft2 x 49.375 ft = 180.52 ft3 3 3 3 Total pipe volume = 481.98 ft + 13.40 ft + 320.81 ft + 13.40 ft + 180.52 ft 3 3 1010.11 ft3 Total horizontal pipe surface area = D x L (Horizontal Length)

=2.158 ftx (131.833 ft+ 3.666 ft+ 87.75 ft+ 3.666 ft+ 49.375 ft)

= 2.158 ft x 276.29 ft = 596.23 ft2 3 Total horizontal pipe volume = 10 10.11 ft 7.3.3.5 Volume & Surface Area for Aerosol Deposition - RPV to Outboard MSIV 035:

Combined total pipe volume V3 RPV Nozzle to Inboard MSIV + Inboard MSIV to Outboard MSIV

= 301.81 ft3.(Section 7.3.3.1) + 101.89 ft3 (Section 7.3.3.2) = 403.70 ft3 Combined total horizontal pipe surface area 2

= 49.89 ft (Section 7.3.3.1) + 66.18 ft (Section 7.3.3.2) = 116.07 2 ft2 Combined total horizontal pipe volume 76.83 ft3 (Section 7.3.3.1) + 101.89 ft3 (Section 7.3.3.2) = 178.72 ft3 7.3.3.6 Volume & Surface Area for Aerosol Deposition - Outboard MSIV 035 to TSV MSV 3 Combined total pipe volume V4 = Outboard MSIV to Third MSIV + Third3MSIV to TSV MSV 3.

3

= 125.89 ft (Section 7.3.3.3) +1,010.11 ft (Section 7.3.3.4) = 1,136.00 3 ft Crombined total horizontal pipe surface area 2

= 33.59 ft2 (Section 7.3.3.3) + 596.23 ft2 (Section 7.3.3.4) = 629.82 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 7.4 Plateout of Activity in Main Steam Lines 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. RGI 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 CC-AA-309-1001, Rev 3

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO. 3 PAGE NO. 42 of 109, 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 40 percentile settling velocity is reasonably conservative for aerosol deposition in the MSIV leakage. The results of the Monte Carlo analysis for settling velocity in the main steam line are given in the following Table:

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

60t' (average) 0.00148 11.43 50th (median) 0.00117 9.04 40tf 0.00081 6.26 1 Oth 0.00021 1.62 The Staff concluded that use of a 10th 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., 50 thpercentile settling velocity) (Ref. 10.22, page 11). This analysis is conservative relative to the Staff's recommendation, in that it models a.30 percentile settling velocity for aerosol deposition in the MSIV leakage paths beyond the outboard 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 *Ciýn-Q* C-Xs*V*C (1) dt Where V volume of well-mixed region C = concentration of nuclides in volume Q volumetric flow rate into volume

= rate constant for settling And u

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:

I CC-AA-309-1001, Rev 3 1

CALCULATION NO. H-l-ZZ-MDC-1880 REVISION NO. 3 PAGE NO. 43 of 109 C =1Ci*

1+ 2.*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:

lftl C =1- 1 (2)

I +

Cim 1+ .* 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 & 100 scfh in the MSIV failed and intact steam lines, respectively.

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 cfm= 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 7.6.1 Iodine/Aerosol Deposition on CREF Charcoal/J-IEPA Filter- Containment Leakage:

Iodine Activity Deposited On CREF Charcoal Filter:

As shown in Figure 5, the CRintake and recirculation charcoal filter elemental & iodine removal efficiency is 99% with the intake and recirculation flow rates of 1,000 cfm and 2,600 cfm, respectively.

Charcoal Filter 1.=99%

1000 'in IOI 990 Iin Suppose Air Intake Iodine Activity (atoms) = Iin and Iodine Activity Deposited on Recirc Filter = Ide Activity deposited on the intake charcoal filter = In x.0.99 x 1,000 cfm = 990 Ii, Filtered inflow activity introduced into the CR = Iin X ( - 0.99) x 1,000 cfm = 10 Ii, CC-AA-309-1001, Rev 3.

I CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO. 3 - PAGENO. 44 of 109 10 Iin + 350 Irn (unfiltered inleakage) = 360 In Recirc Charcoal Filter 11= 99% 3.6 Tin Ide.

356.4 i Activity deposited on recirc charcoal filter (conservatively assuming only one pass through. the recirc

'charcoal filter) Ide = 0.99 x (360 Iin cfmn) = 356.4 Tin; Rearranging: Tin Ide, 356.4 Therefore, the activity deposited on the intake charcoal filter = 990 Tin= (990/356.4) Ide Total elemental & organic iodine activity deposited on intake + recirc charcoal filter

= (990/356.4 Ide) + Ide,: 3.78 Ide Ide = 9.51,76E+13 Atoms (Elemental Iodine) + 3.1122E+14 Atoms (Organic Iodine) (HEPU350CL02.o0, CR Recirculating Filter Nuclide Inventory @ 720 hrs) 4.06396E+14 Atoms Total iodine activity deposited on the CR intake + recirc charcoal filter due to containment leakage

= 3.78 x 4.06396E+14 Atoms = 1.5362E+15 Atoms Aerosol Mass Denosited On CREF 1-IEPA 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 1,000 cfm and 2,600 cfm respectively.

1000 Ain 10 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 1,000 cfm = 990 Ain Filtered inflow aerosol mass introduced into the CR ý Ain x (I - 0.99) x 1,000 cfm= 10 Ain 10 Ain + 350 Ain (unfiltered inleakage) = 360 Ain I

Recirc HEPA Filter 3.6 Ain 71= 99%, v, Ade 356.4 Ain Aerosol mass deposited on recirc HEPA filter (conservatively assuming only one pass through the recirc HEPA filter) = Ad, 0.99 x (360 Ain cfm) = 356.4 Ain; Rearranging: Ain = Ade / 356.4 CC-AA-309-100 1, Rev 3

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO. 3 PAGE NO. 45 of 109 Therefore, the aerosol mass deposited on intake HEPA filter = 990 Ain= (990/316.8) Ade Total aerosol mass deposited on intake + recirc HEPA filter

= (990/356.4 Ade) +/- Ad, t 3.78 Ade Ad, = 8.8810E-09 kg (HEPU350CL02.oO, CR Recirculating Filter Nuclide Inventory @ 720 hrs)

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

= 3.78 x 8.8810 E-09 kg = 3.3570E-08 kg 7.6.2 Iodine/Aerosol Deposition on CREF Charcoal/HEPA Filter -ESF Leakage:.

Iodine Activity Deposited On CREF Charcoal Filter: I As discussed in Section 7.6.1 above:

Total elemental & organic iodine activity (atoms) deposited on intake + recirc charcoal filter 3.78 Id, Ide = 7.5326E+16 Atoms (Elemental Iodine) + 2.3297E+15 Atoms (Organic Iodine) (HEPU350ES02.oO, CR Recirculating Filter Nuclide Inventory @ 720 hrs)

= 7.7656E+16 Atoms.

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

= 3.78 x 7.7656E+16 Atoms = 2.9354E+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 t recirc HEPA filter (HEPU350ES02.oO, CR Compartment Nuclide Inventory @ 720 hrs).

7.6.3 Iodine/Aerosol Deposition on CREF Charcoal/HEPA Filter - MSIV Leakage:

Iodine Activity Deposited On CREF Charcoal Filter:

As discussed in Section 7.6,1 above:

Total elemental& organic iodine activity (atoms) deposited on intake-+ recirc charcoal filter = 3.78 Id, Ide 3.3511E+14 Atoms (Elemental Iodine) + 1.8198E+15 Atoms (Organic Iodine)

(HEPU350MS02.o0, CR Recirculating Filter Nuclide Inventory @ 720 hrs)

= 2.1549E+15 Atoms.

Total iodine atoms deposited on the CREF intake + recirc charcoal filter due to the MSIV leakage 3.78 x 2.1549E+15 Atoms = 8.1455E+15 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 3.78 Ade Ad, = 4.2677E-09 kg (HEPU350MS02.oO, CR Recirculating Filter Nuclide Inventory @ 720 hrs)

Total aerosol mass deposited on CREF intake + recirc HEPA filter due to the MSIV leakage

= 3.78 x 4.2677E-09 kg = 1.6132E-08 kg CC-AA-309-100I, Rev 3

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO.3 PAGE NO.46 of1 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 1.5362E+15 Atoms + 2.9354E+17 Atoms + 8.1455E+15 Atoms = 3.0322E+17 Atoms, which is used in Table 16 to obtain the total iodine isotopic activities on the CREF charcoal.

Total aerosol mass deposited on the CREF HEPA Filter

= Aerosol mass from containment leakage + Aerosol mass from ESF leakage + Aerosol mass from MSIV leakage

=3.3570E-08 kg + 0.00 kg + 1.6132E-08 kg = 4.9702E-08 kg, which isused in Table 18 to obtainthe total aerosol isotopic activities on the CREF HEPA Filter.

7.6.4 CREF Shieldina 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:

6"Concrete Pad CREFCE W "."

4 -- 2 -279-2 "

'[*1 C14.25' 27.52' 27.386\

4--27.52 -

CR Console Elevation View Plan View Horizontal distance [(27'-2",)2 + (4'-4-1/2") 2 = 27.52' CREF center line elevation

= CR ceiling elevation + Thickness of concrete base + 12 CREF height [see Figure 5]

= 155'-3" (Ref. 10.39.c) + 6" (Ref. 10.39.c) + V2 (3'-0".) =157'-3" 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,)2I1]/ = 30.99' Slant Angle Through Concrete Ceiling = 0 = 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' I CC-AA-309-1001, Rev 3

CALCULATION NO. i1-I-ZZ-MDC-1880 REVISION NO. 3 PAGE NO. 47 of 109 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 3.966E-02 mRem/hr (MicroShield Run HCCRFLT2.MS5)

CR Operator Exposure Time = xx (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

= 3.966E-02 mRem/hr x 1/1000 Rern/mRem x 316.8 hr 1.26E-02 Rem 7.7 FRVS Vent & Recirc, and CR Charcoal/HEPA Filters Efficiencies REPA 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)

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% - rl)/safety factor (100% - -9)/2 Where rl = HEPA filter efficiency to be credited in the analysis 0.05%= (100%-rl)/2 0.1%= (100%- i) rl= 100% - 0.1% = 99.9%

Conservatively, the BEPA 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% - r)/safety factor = (100%- -j)/2 Where r1 = CREF charcoal filter efficiency to be credited in the analysis CFREF Charcoal Filter 0.5%.= (100% -,n)/2 1%= (lO0 %-i)~

rI 100%- 1% = 99%

FRVS Vent Charcoal Filter Elemental Iodine r1= 90% (Ref. 10.47)

Organic Iodine rj = 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 I .CC-AA-309-1001, Rev 3

CALCULATION NO, H-1-ZZ-MDC-1880 REVISION NO.3 PAGE NO. 48 of 109 7.7 Isotopic Activities Released To Environment The isotopic activities released to the environment at various post-LOCA time intervals are listed in Tables 20 through 24. These isotopic activities are obtained from the RADTRAD computer Runs JEPU350CLO2.o0, HEPU350ES02.o0, and HEPU350MS02.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 Hiaher 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 Reference 10.24. The use of a smaller wetted surface is conservative because it results in a smaller elemental removal coefficient and it takes a longer time to reach an elemental iodine decontamination factor (DF) of 200, which allows the elemental iodine to remain airborne in the drywell atmosphere for release to the atmosphere Via containment and MSIV leakage.

Total drywell surface area 1.52,261 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, whichare 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 ft2 (17,850 ft2 /4 4,463 ft2)

Downcomer (to water level) = 3,168 ft2 Vent header & line = 9,727 ft2 Suppression chamber (to water level) 15,408 ft2.

Estimated 50% of major equipment = 5,306 ft22 (10,612 ft2 /2 = 5,306 ft22 )

Estimated 50% of structures = 6,181. ft (12,361 ft2/2 = 6,181 ft )

Total estimated wetted drywell surface area = 44,253 ft2 (69,126 fta without reductions) 75% of total estimated wetted drywell surface area . 33,200 ft2 (0.75 x 44,253 ft2) 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:

w= Kw x A/V Where:

W= first order removal coefficient by wall deposition K,* =mass transfer coefficient = 4.9 m/hr (Ref. 10.41, page 6.5.2-10)

CC-AA-309-I001, Rev 3

CALCULATION NO. H-1-ZZ-MDC-1 880 REVISION NO.3 PAGE NO.49 of 109 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 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 Xw = K, x A/V = 4.9 rn/hr x (3.2808 ft/rm) (33,200 ft2) / (3.06E+05 ft3) = 1.74 hf1 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.

7.12 Steam Mass Flow Rate:

Uprated Steam Flow Rate

= 17,774,000 lb/hr (Ref. 10.69, Section 3.2.1) x 45 3 .6 g/lb x 1/3600 hr/sec

= 2,239,524.0 g/sec - 2.240E+06 g/sec This conversion factor is used in Table .12 to convert thenoble gas release rates in ptCi/sec to noble gas activity concentrations in 4Ci/g.

7.13 Determination of Reactor Coolant Volume & Mass:

Reactor Coolant Volume Based on HC UFSAR Table 6.2-3 (Ref. 10.63)

RCS volume of liquid in vessel 11,885 ft3 RCS volume of liquid in recirculation loops = 1,214 ft3 Total RCS liquid in vessel & recirculation loop = 11,885 ft3 + 1,214 ft3 = i3,099 ft3 RCS volume of steam in vessel = 9,089 ft3 Total water and steam volume of reactor vessel & RCS = 21,970 ft3 (Ref. 10.6.22)

Net water volume of reactor vessel & RCS

= Total water and steam volume of reactor vessel & RCS - RCS volume of steam in vessel

= 21,970 ft3 - 9,089 ft3 = 12,881 ft3, which is less than 13,099 ft2 determined based on the UFSAR Table 6.2-3, above, therefore, the RCS coolant volume of 13,099 ft3 is conservatively used in the analysis to maximize the RCS activity available for release to the environment.

Total RCS coolant volume conservatively maximized by being based on water density of 62.4 lb/ft3 at standard temperature and pressure conditions rather than a lower water density at operating conditions.

= 13,099 ft3 x 62.4 lb/ft3 x 453.6 gm/lb = 3.708E+08 gm used in Table 15.

CC-AA-309-1001, Rev 3 1

CALCULATION NO. H-I-ZZ-MDC-1880 TREVISION NO. 3 PAGE NO. 50 of 109 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 4.79E-01 3.87E-0 I 1.31E-01 (311 hr)

RCS Activity Release 1.88E-02 2.15E-02 21.15E-03 Via Open PCIV (0.0 hr)

ESF Leakage 3.33E+00 5.22E-01 2.64E-01 (14.6 hr)

MSIV Leakage 3.31E-01 5.01E-01 1.51E-01

~(4.7 hr)

Containment Purge O.OOE+00 O.OOE+00 O.OOE+00 Containment Shine O.OOE+00 O.OOE+00 O.OOE+00 External Cloud O.OOE+00 O.OOE+00 O.OOE+00 CR Filter Shine .1.26E-02 O.OOE+00 O.OOE+00 Total 4.17E+00 1.43E+00 5.48E-01

.Allowable TEDE Limit 5.OE+00 2.50E+01 2.50E+01 RADTRAD Computer Run No.

Containment Leakage HEPU350CL02.o0 HEPU350CLO2.oO IEPU350CLO2.oO RCS Activity Release HEPURCSCL00.o0 HEPURCSCLOO.oO HEPURCSCLOO.oO ESF Leakage HEPU350ES02.oO HEPU350ES02.oO HEPU350ES02.oO MSIV Leakage HEPU350MS02'.oO HEPU350MS02.oO HEPU350MS02.o0 CC-AA-309-1001, Rev 3

CALCULATION NO. H-1-ZZ-MDC-1880 I REVISION NO. 3 1 PAGE NO. 51 of 109 8.2 Compliance of revised dose increases with the 10 CFR 50.59 rule is shown in the following table:

Current Proposed Regulatory Proposed Minimum RG Design Basis Accident Total Total Dose Dose Dose Dose Dose Dose Limit Increase Increase 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 3 (LOCA) 1880, Rev 2 Control Room 4.16 4.17 5 0.01 0.084 5 Exclusion Area Boundary 3.10 1.43 25 -1.67 2.19 25 Low Population Zone 0.696 0.548 25 -0.15 2.43 25 C From 10 CFR 50.67 (Ref. 10.4)

F From RG 1.183, Table 6 (Ref. 10.1)

9.0 CONCLUSION

S The Section 8.1 results, based on the revised analysis, which: uses the latest aerosol deposition mechanism approved by the NRC staff in other successful AST license.amendments for the MSIV leakage release paths, and the reduced ESF leakage flashing fraction developed based on the plant-specific design bases, 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 for the maximum PCIV isolation time up to 120 seconds.

The time between accident initiation and the onset of the post-LOCA gap release phase is 120 seconds (Ref. 10.1, Table 4). Therefore, no core gap activity is released in the containment while the PCIVs remain open for 120 seconds. The primary coolant activity released during this 120 seconds time period will not add to the 180-day EQ integrated dose. Therefore, the changes in this calculation will not impact the EQ doses calculated in H-1-ZZ-MDC-1931, Rev 0.

The comparisons in Section 8.2 confirm that the proposed increase in the CR total calculated dose is less than the minimal dose increase margin and allowable regulatory limit. The EAB and LPZ doses are less than the previously calculated values.

I CC-AA-309-1001, Rev 3

CALCULATION NO. H-1-ZZ-MDC-188F REVISION NO.3 PAGE NO.52 of 109

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 S1. Humphreys et al., "RADTRAD V3.02: A Simplified Model for Radionuclide Transport and Removal and Dose Estimation," NUREG/CR-6604, USNRC, April 1998.

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

10.4 10 CFR 50.67, "Accident Source Term."

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

10.6 HCGS Technical Specifications:

10.6.1 Specification 4.6.5.3.1 .c. 1, FRVS Vent HEPA Filter Testing Criterion 10.6.2 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 Thtermal Power.

10.6.10 Specification 4.6.5.3.2.c. 1, FRVS Recirc HEPA Filter Testing Criterion

.10.6.11 Specification 3/4.4.5, "Specific Activity" Limiting Condition for Operation 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.c.1, Control Room Emergency Filtration System Surveillance Requirements 10.6.14 Specification 4.7.2.c.2, Control Room Emergency Filtration System Surveillance Requirements 10.6.15 Specification 4.7.2.c.3, Control Room Emergency Filtration System Surveillance Requirements 10.6.16 Specification 4.7.2.e.3, Control Room Emergency Filtration System Surveillance Requirements 10.6.17 Specification 3.6.1.2.c, Primary Containment Leakage Limiting Condition For Operation 10.6.18 Specification 3.6.1.8, Drywell.and Suppression Chamber Purge System 10.6.19 Not Used 10.6.20 Specification 1.11, Dose Equivalent 1-13 1.

CC-AA-309-1001, Rev 3

CALCULATION NO. H-1-ZZ-MDC-1880 REVISI1ON NO. 3 P.AGE NO. 53 of 109 10.6.21 Specification 1.12, E-Average Disintegration Energy.

10.6.22 Specification 5.4.2, Reactor Coolant System Volume 10.7 Federal Guidance Report 11, EPA-520/1-88-020, Environmental Protection Agency.

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

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

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

10.11 Drawing No. I-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-0 11, Rev 11 10.121f .- 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-1414-1 Rev 1,Reactor Building Area 14, Section D14 - D14 10.13.h Pm13044I, Rev 4, Piping Area Drawing, ReactQr Building Area_13, Plan at El. 132'-0"

& El. 145' 10.13.i P-1711-1, Rev 2, Reactor Building Area 17, Section A17 -A17 10.14 Piping Class Sheet Drawing No. 10855-P-0500:

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

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

10.17 Vendor Technical Document (VTD) No. 430024, Volume 002, EPU TR T0400 - Containment System Response 10.18 Not Used.

CC-AA-309-1001, Rev 3

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO. 3 PAGE NO. 54 of 109 10.19 Calculation No. GU-00 13, Rev. 4, Filtration Recirculation and Ventilation System Exhaust Rate 10.20 Not Used.

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 Not Used.

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 EQEInternational, 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" and 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-13.17-0, Rev 22, Floor Plan EL 155'-3" Area 25 I0.29.b C-1319-0, Rev 12, Floor Plan EL 155'-3" Area 26 16.29.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-I-ZZ-MDC-l1882, Rev 0, Control Room Envelope Volume.

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

CC-AA-309-1001, Rev.3

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO. 3 PAGE NO. 55 of 109 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.

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 PSE&G Vendor Technical Document (VTD) No. 325236, Control Room Envelop Inleakage Testing At Hope Creek Generating Station, Final Report; 2001.

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

Subject:

FRVS Vent Charcoal Filter Efficiencies (Attachment A) 10.48 Critical Software Package Identification No. A-0-ZZ-MCS-0225, Rev 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 Not Used.

10.51 Not Used.

10.52 Not Used.

10.53 Not Used.

10.54 NRC letter to PSEG Nuclear dated October 3, 2001, "Hope Creek Generating Station - Issuance of Amendment Re: Increase In Allowable Main Steam Isolation Valve (MSIV) Leakage Rate and Elimination of MSIV Sealing System (TAC No. MB 1970)."

10.55 Not Used.

10.56 Not Used.

10.57 Not Used.

CC-AA-309-1001, Rev 3

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO. 3 PAGE NO. 56 of 109 10.58 Not Used.

10.59 Hope Creek Operating License Amendment No. 171, RE: Relocate Component Lists For Primary Containment Isolation Valves From Technical Specifications (TAC No. MD3600).

10.60 Hope Creek Generating Station Technical Requirements Manual (HC TRM), Revision 1, Table 3.6.3-1, Primary Containment Isolation Valves.

10.61 Hope Creek P&IDs:

10.61.1 M-13-1, Rev 37, Reactor Auxiliaries Cooling 10.61.2 M-25-1, SHT 1, Rev 17, Plant Leak Detection 10.61.3 M-41-1, SHT 1, Rev 35, Nuclear Boiler 10.61.4 M-43-1, SHT 1, Rev 33, Reactor Recirculation System 10.61.5 M-44-1, SHT 1, Rev 31, Reactor Water Claenup 10.61.6 M-51-1, SHT 1, Rev 37, Residual Heat Removal 10.61.7 M-51-1, SHT 2, Rev 35, Residual Heat Removal 10.61.8 M-52-1, Rev 30, Core Spray 10.61.9 M-53-1, SHT 2, Rev 26, Fuel Pool Cleaning & Torus Water Clean-up 10.61.10 M 1, SHT 1, Rev 40, ContainmentAtmosphere Control 10.61.11 M-58-1, Rev 09, Containment Hydrogen Recombination System 10.61.12 M-59-1, SHT 1, Rev 31, Primary Containment Instrument Gas 10.61.13 M-59-1, SHT 3, Rev 6, Primary Containment Instrument Gas 10.61.14 M-61-1, SHT 1, Rev 24, Liquid Radwaste Collection 10.61.15 M-61-1, SHT 2, Rev 20, Liquid Radwaste Collection 10.61.16 M 1, Rev 19, Reactor BuildingAir Flow Diagram 10.61.17 M-87-1, SHT 3, Rev 24, Chilled Water System, Turbine Building Chilled Water 10.61.18 M-87-1, SHT.2, Rev 23, Chilled Water System, Reactor Building & Drywell.Chilled Water 10.62 Not Used.

10.63 Hope Creek:UFSAR Table 6.2-3, Initial Conditions for Containment Response Analyses.

10.64 Not Used 10.65 A list of the I&C Primary Containment Isolation Valve Testing Procedures affected by the increased valve closure time is provided by Hope Valve Program Management, which is included as an affected operating procedures due to the proposed activity. This list shown in Attachment 14.3, which requires an independent verification from the Hope Creek Mechanical Engineering.

10.66 Technical Information Document (TID)-14844, Calculation of Distance Factors for Power and Test Reactor Sites, March 23, 1962.

10.67 NUREG-0800, Standard Review Plan, "Radiological Consequence Analyses Using Alternative Source Terms," SRP 15.0.1, Rev. 0, July 2000 CC-AA-309-1001, Rev 3

I REVISION NO.3 I CALCULATION NO. H-I-ZZ-MDC-1880 I PAGE NO. 57 of 109 10.68 Vendor Technical Document (VTD) No. PNO-A61-4100-0047, Rev 2, GE Specification Document No. 22A2703F, Rev 3, Radiation Sources 10.69 Vendor Technical Document (VTD) No. 430059, Volume 002, EPU TR T0807 - Coolant Radiation Sources 10.70 Dresden, Units 2 & 3, & Quad Cities, Units 1 & 2, License Amendment, Issuance of Amendments Re: Adoption of Alternative Source Term Methodology, MB6530, MB653 1, MB6532, MB6533, MC8275, MC8276, MC8277 & MC8278, September 11, 2006 (ADAMS Accession #ML062070290) 10.71 Peach Bottom, Units 2 and 3 - Issuance of Amendments 269 and 273 Re: Application of Alternative Source Term Methodology, September 5, 2008 (ADAMS Accession

ML082320406) 10.72 Vermont Yankee Nuclear Power Station - Issuance Of Amendment Re: Alternative Source Term (TAC No. MC0253), March 29, 2005 (ADAMS Accession #ML041280490)

CC-AA-309-1001, Rev 3

CALCULATION NO. H-!-ZZ-MDC-1880 REVISION NO.3 PAGE NO. 8 of 109 11.0 TABLES Table 1A Higher Core Inventory Including Parent/Daughter Isotopes Per RADTRAD Table 1.4.3.3-2 Discharge Bundle Core Core Core Core Isotope. Inventory Inventory Isotope Inventory Inventory (CiMWt) (Ci/MWt) (Ci/MWt) (Ci/MWt)

CO-58* 1.529E+02 1.529E+02 E 4-6 CO-60* 1.830E+02 1.830E+02 KR-85 4.711E+02 4.71IE+02 TE-132 3.917E+04 3.917E+04 KR-85M 5.908E+03 5.908E+03 1-131 2.7.79E+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"

SR-90 3.790E+03 3.790E+03 47- q A I4ý XE-133 5.306E+04 5.306E+04

-,2 )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 LS 77220

Y-92 3.005E+04 3.005E+04 i S32.E*Q3 "

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

Ij LA-140 4.981E4 4. E+04 NT'l 7"%1 1 1EF!1-! LA- 141 4.325E+04 4.325E+04 N1 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 5.278E4-04 CE-143 3-910E+04 3.910E+04 TC-99M 4.621E+04 4.621E+04 8P RU-105 3.529E+04 3.529E+04 PR-143 3.783E+04 3.783E+04

- ND-147 1.783E+04 1.783E+04 C, NP-239 6.917E+05 6.917E+05 RH-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 1.333E+01 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 I &O TE-129M ao 1-401E+03a 1.401E+03 CM-244 u5.Us E+02 [ 5.1 88E+02

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

CC-AA-309-1001, Rev 3 1

CALCULATION NO. H-,-ZZ-MDC-1880 REVISION NO.3 PAGE NO.59 of 109 Table 1B Core Inventory - Discharge Bundle Exposure Total Total Core Core Isotope Inventory Isotope Inventory (CiJMWt) (Cj/MfWt)

CO-58* 1.529E+02 TE-131M 2.876E+04 CO-60* 1.830E+02 TE-132 3,917E+04 KR-85 4.71 1E+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 4.231E+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 ZRM97 1.307E+05 LA-142 4.134E+04 NB-95 4.237E+04 CE-141 4.350E+04 MO-99 5.278E+04 CE-143 3.910E+04 TC-99M 4.621E+04 CE-144 7.234E+04 RU-103 8.941E+04 PR-143 3.783E+04 RU-105 3.529E+04 ND-147 1.783E+04 RU-106 4.722E+04 NP-239 6.917E+05 RH-105 3.237E+04 PU-238 3.442E+02 SB-127 3.379E+03 PU-239 1.333E+01 5B-429 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 IA CC-AA-309-100 1, Rev 3

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO.3 PAGE No. 60 of 109 Table IC Higher Core Inventory Including Parent/Daughter Isotopes Per RADTRAD Table 1.4.3.3-2 Core Avera e Exposure Core Core Core Core Inventory Inventory Inventory Inventory Isotope Isotope (Ci/MWt) (CiIMWt) (Ci/MWt) (Ci/mwt)

CO-58* 1.529E+02 1.529E+02 r.-* 6 i C0~60* 1.830E+02 1.830E+02 4EQ7 iuE#QI KR-85 3.300E+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 >* . ............ O 2 4

SR-90 2.640E+03 2.640E+03 *XB>W,.0.,Q . .......... __

C*2 XE-133 5.300E+04 5.300E+04 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 2.810E+03 2.810E+03 CS-136 1.860E+03 1.860E+03 Y-9 1 3.440E+04 3.440E+04 L 740F SY-92 3.620E+04 3.62013+04 B -:

Y-93 4.160E+04 4.160E+04 BA-139 4.950E+04 4.950E+04 ZR-95 41850E+04 4.850E+04 BA-140 4.780E+04 4.780E+04

LA-140 5.080E+04 5.080E+04 N- t LA- 141 4.510E+04 4.540E+01 T,16 LA-142 4.370E+04 4.370E+04 NB-95 4.870E+04 4.870E+04 CE-141 4.540E+04 4.540E+04 MO-99 5.100E+04 5. 100E+04 CE-143 4.220E+044.2E4 TC-99M 4,460F+04 4.460E+04 * (143* * ~ N SB-19 8490+03 8.40E+0 PU240 1.40E+1 14.10E+t04 RU- 105- 2.70E+04 2.700E+014 PR- 143 4.080E+04 4.08013+04

....,..... ~.10

... .,* }* * -ij0 - W... N D -147 1 .81OE .810E + 04

- [*

S}*NP-239 5.220E,+05 5.220E+05 RH- 105 2.530E+04 2.530E+04 PU-238 9.040E+01 9.040E+01 SB-127 2.800E+03 2.800E+03 PTU-239 1.09013+01 1.090E+01 SB-129 8.490E+03 8.490E+03 PU-240 1.410E+01 1.410E+01 TE-127 2.780E+03 2.780E+03 PU-241 4.090E+03 4.090E+03 TE-127M 3.7 10E+02 3.710E+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-129M 1.240E+03 1.240E+03 CM-244 5.240E+01 5.240E+O1

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

I CC-AA-309-1001, Rev 3

I CALCULATION NO. H-1-ZZ-MDC-1880 I REVISION NO.3 PAGE NO.61 of 109 Table 1D Core Inventory - Core Average Exposure Total Total Core Core Isotope Inventory Isotope Inventory (Ci/Mwt) (CVMWt)

CO-58* 1.529E+02 TE-131M 2.764E+04 CO-60* 1.830E+02 TE-132 3.810E+04 KR-85 3.300E+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 4.5 1OE+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 21800E+03 PU-239 1.090E+01 SB-129 8.490E+03 PU-240 1.410E+01 TE 127M 2.780E+03 PU-241 4.090E+03 TE-127 3.710E+02 AM-241 4.600E+00 TE-129M 8.350E+03 CM-242 1.090E+03 TE-129 1.240E+03 CM-244 5.240E+OI Total Core Inventory From Table IC CC-AA-309-1001, Rev 3

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO.3 PAGE NO. 62 of 109 Table 2 Hope Creek Main Steam Horizontal Piping Volume & Surface Area Table 3 Hope Creek Main Steam Piping Volume Main Steam Piping Inside Volume (ft3)

Header Header A D Piping Between RPV Nozzle & 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 I CC-AA-309-1001, Rev 3 1

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO. 3 PAGE NO.63 of 109 Table 4 Main Steam Piping Parameters Used In MSIV Leakage Release Path Model Main Steam Piping Between RPV Nozzle & Outboard Outboard MSIV &

Main Steam MSIV Turbine Stop Valve Header Horizontal Horizontal ID Inside Surface Inside Surface Volume Area Volume Area (ft3) (ft2 ) (ft) (ft)

MSIV Failed Line Steam HeaderA* 101.89 66.18 1065.25 631.62 (First Shortest Line)._

MSIV Intact Line Steam Header D 178.72 116.07 1062.21 629.82 (Second Shortest Line)

Main Steam Header Volume and Surface Area Information From Table 2

Information Pipe Segment Between Inboard & Outboard MSIVs Table 5 Rate Constant for MSIV Leakage Release Path with 50%/3.0% Settling Velocities.

Settling Horizontal Horizontal Rate Peach Bottom Velocity Settling Pipe Constant for Steam Header Area Volume Settling (ft/h r) (ft2) (ftz) (hrq)

A B C D MSIV Failed Line - Header A Inboard MSIV To 13.82 66.18 101.89 8.98 Outboard MSIV MSIV Failed Line - Header A Outboard MSIV To 7.08 631.62 1065.25 4.20 Turbine Stop Valve SV-3 MSIV Intact Line - Header B RPV Nozzle To 13.82 116.07 178.72 8.98 Outboard MSIV MSIV Intact Line - Header B Outboard MSIV To 7.08 629.82 1062.21 4.20 Turbine Stop Valve IIII A = 50 Percentile Settling Velocity = 0.00117 m/sec x 3.28 ft/m x 3600 sec/hr = 13.82 fthr for main steam lines upstream of outboard MSIVs i

A 30 Percentile Settling Velocity = 0.0006 m/sec x 3.28 ft/m x 3600 sec/hr = 7.08 ft/hr for main steam lines downstrearn of outboard MSIVs B & C From Table44 D = .ý = (Ax B)/C CC-AA-309-1001, Rev 3

CALCULATION NO. H-,-ZZMDC-1880 REVISION NO. 3 PAGE NO. 64 of .109 Table 6 Gravitational Deposition Aerosol Removal Efficiency On Horizontal Pipe Surface With 50%/30% Settling Velocity (250 scfh)

Post-LOCA Settling Well Volumetric ' Aerosol Post-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 V1 MSIV Failed V3 Intact A B Line A B Line (hr) (hr-') (ft3 ) (fta/hr) (%) (hr) (hr-') (ft 3 ) (ft3 /hr) (%)

MSIV Failed Main Steam Line Between Inboard & Outboard MSIVs Intact Main Steam Line Between RPV & Outboard MSIV 0-24 8.98 101.89 150.00 85.92 0-24 8.98 403.70 100.00 97.32 24-96 8.98 101.89 75.00 92,42 24-96 8.98 403.70 50.00 98.64 96-720 8.98 101.89 75.00 0.00 96-720 8.98 403.70 50.00 0.00 Post-LOCA Settling Well Volumetric Aerosol Post-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 X,. V2 MSIV Failed V4 Intact A B Line A B Line (hr) (hr-') (ftJ) (ft3/hr) (%hOr) (he-*) (fta) (ft/h r) %

MSIV Failed Main Steam Line Between Outboard MSIV & TSV Intact Main Steam Line Between Outboard MSIV & TSV 0-24 4.20 1139.04 150.00 96.96 0-24 4.20 1136.00 100.00 .97.95 24-96 4.20 1139.04 75.00 98.46 24-96 4.20 1136.00 50.00 98.96 96-720 4.20 1139.04 " 75.00 0.00 96-720 4.20 1136.00 50.00 0.00 A From Table 5 B From Table 3 Intact Main Steam Line Between RPV & Outboard MSIV = 301.81 ft' + 101.89 ft3 = 403.70 ft3 The aerosol removal is not credited in any steam line 96 hrs after the onset of a LOCA CC-AA-309-1001, Rev 3

CALCULATION NO. R-1-ZZ-.MDC-1880 REVISION NO. 3 PAGE NO. 65 of 109 Table 7 MSIV Leak Rate In Different Control Volume (Total = 250 scfh & Max = 150 scfh)

MSIV Leak Rate In Various Control Volumes (efh)/(cfm)

Post-LOCA Drywell To Volume V1 VolumeV 2 Drywell To Intact Line Volume V4 Time MSIV Failed To To Intact Line Volume V3 To Interval Volume V1 Volume V2 Atmosphere Volume V3 To Atmosphere (hr) ,_, _Volume V4 48.47 150.00 150.00 32.31 100.00 100.00 0-2

0.808 2.500 2.500 0.539 1.667 1.667 26.77 150.00 150.00 17.84 100.00 100.00 2-24 0.446 2.500 2.500 0.297 1.667 1.667 13.39 75.00 75.00 8.92 50.00 50.00 0.223 1.250 1.250 0.149 0.833 0.833 MSIV Leak Rate Information From Section 7.2 CC-AA-309-1001, Rev 3

CALCULATION NO. H-1-ZZ-MDC-1880 I REVISION NO.3 PAGE NO. 66 of 109 i 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/MB 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 & 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 5.51OE+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 1D B From Section 7.9 D Reference 10.1, Appendix A, Section 3.3 E From table 8 CC-AA-309-1001, Rev 3 1

CALCULATION NO. H-1-ZZ-MDC-1880 RE VISION NO.3 PAGE NO.67 of 109 Table 10 Scaling Factor for Specific Iodine Concentration Normal Iodine Iodine Dose Product Isotope Activity Conversion Concentration Factor jiCi.rem/Ci.g Cui/g (rem/Ci) (rem)

A B (A x B)

I- 131 1.300E-02 1.48E+06 1.924E+04 1-132 1.200E-01 5.35E+04 6.420E+03 1-133 8.900E-02 4.OOE+05 3.560E+04 1-134 2.400E-01 2.50E+04 6.OOOE+03 1-135 1.300E-01 1.24E+05 1.612E+04 Total 8.338E+04 A From Reference 10.69, Appendix A B From Reference 10.66 1-131 DE Based on Normal Iodine Concentration 5.634E-02 jIodine Scaling Factor Based on 0.2 gCi/g DE 1-131 3.550E+00 Table 11 Iodine Concentration Based On Specific Iodine Concentration Normal Iodine Iodine Iodine Scaling Activity Isotope Activity Factor Concentration Concentration RtCi/g PtCi/g A B C=AxB 1-131 1.300E-02 3.550E+00 4.615E-02 1-132 1.200E-01 3.550E+00 4.260E-01 1-133 8.900E-02 3.550E+00 3.160E-01 1-134 2.400E-01 3.550E+00 8.520E-01 1-135 1.300E-01 3.550E+00 4.615E-01 A From Reference 10.69, Appendix A B Scaling Factor Based on 0.2 pCi/g DE 1-131 From Tablel10 I CC-AA-309-1001, Rev 3 1

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO.3 PAGE NO.68 of 109 Table 12 Normal Noble Gas Concentration Noble Gas Steam Mass Normal Release Rate Mass Noble Gas Isotope At t =0 Flow Rate Activity

(ýtCi/sec) (g/sec) Concentration (OC/g)

A B C=A/B Kr-83m 3.400E+03 2.240E+06 1.5 18E-03 Kr-85m 6.100E+03 2.240E+06 2.723E-03 Kr-85 2.OoOE+01I 2.240E+06 8.929E-06 Kr-87 2.OOOE+04 2.240E+06 8.929E-03 Kr-88 2.OOOE+04 2.240E+06 8.929E-03 Xe-131m 1.500E+01 2.240E+06 6.696E-06 Xe-133m 2.900E+02 2.240E+06 1.295E-04 Xe-133 8.200E+03 2.240E+06 3.661E-03 Xe-135m 2.600E+04' 2.240E+06 1.161E-02 Xe-135 2.200E+04 2.240E+06 9.821E-03.

Xe- 138 8.900E'+04 2.240E+06 3.973E-02 A From Reference 10.68, Table V B From Section 7.12 CC-AA-309-1001, Rev 3

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO. 3 PAGE NO. 69 of 109 Table 13 Reactor Coolant Concentration Based on 100/E-BAR Normal Average Energy Weighted EPU Mev/Dis -Energy Isotope Activity Beta Gamma Total E-Bar Concentration Mev.g.Ci/dis.g g+/-Ci/g Ai Bi Ci Di Bi+ Ci Ei = Ai

Di Br-83 1.500E-02 0.321 0.008 0.329 0.0049 Br-84 2.700E-02 1229 1.788 3.017 0.0815 Kr-83m 1.518E-03 0.039 0.003 0.042 0.0001 Kr-85m 2.723E-03 0.255 0.158 0,413 0.0011 Kr-85 8.929E-06 0.251 0.002 0.253 0.0000 KR 87 8.929E-03 1.324 0.793 2.117 0.0189 KR 88 8.929E-03 0.364 1.955 2.319 0.0207 Xe-131m 6.696E-06 0.144 0.020 0.164 0.0000 Xe-133m 1.295E-04 0.192 0.041 0.233 0.0000 Xe-133 3.661E-03 0.136 0.046 0.182 0.0007 Xe-135m 1.161E-02 0.098 0.429 0.527 0.0061 Xe-135 9.821E-03 0.317 0.249 0.566 0.0056 Sr-89 3.1OOE-03 0.583 0.000 0.583 0.0018 Sr-90 2.300E-04 0.196 0.000 0.196 0.0000 Sr-91 6.900E-02 0.656 0.697 1.353 0.0934 Sr-92 1.100E-01 0.196 1.339 .1.535 0.1689 Zr-95 4.OOOE-05 0.116 0.739 0.855 0.0000 Zr-97 3.200E-05 0.700 0.179 0.879 0,0000 Nb-95 4.200E-05 0.044 0.766 0.810 0.0000 Mo-99 2.200E-02 0.392 0.150 0.542 0.0119 Tc-99m 2.800E-01 0.016 0.126 0.142 0.0396 Ru-103 1.900E-05 0.075 0.469 0.544, 0.0000 RudO06. 2.600E-06 0.010 0.000 0.010 0.0000 Te-129m 4.OOOE-05 0.260 0.038 0.298 0.0000 Te-132 4.900E-02 0.102 0.234 0.336 0.0165 Cs-134 1.600E-04 0.164 1.555 1.719 0.0003 Cs-136 .1003E-04 0.139 2.166 2.305 0.0003 Cs-137 2.400E-04 0.187 0.000 0.187 0.0000 Cs-138 1.900E-01 1.207 2.361 3.568 0.6779 Ba-139 1.600E-01 0.898 0,043 0.941 0.1506 Ba-140 9.OOOE-03 0.313 0.183 0.496 0.0045 CC-AA-309-1001, Rev 3

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO.3 PAGE NO.70 of 109 Table 13 (Cont'd)

Reactor Coolant Concentration Based on 100/E-BAR Normal Average Energy Weighted EPU Mev/Dis Energy Isotope Activity Beta Gamma Total E-Bar Concentration Mev.pACi/dis.g pCi/g Ai Bi Ci Di =Bi + Ci Ei =Ai

Di Ba-141 1.700E-01 0.901 0.845 1.746 0.2968 Ce-141 3.900E-05 0.171 0.076 0.247 *0.0000 Ce-143 3.500E-05 0.433 0.282 0.715 0.0000

.Ce- 144 3.500E-05 0.092 0.021 0.113 0.0000 Pr-143 3.800E-05 0.314 0.000 0.314 0.0000 Nd-147 1.400E-05 0.270 0.140 0.410 0.0000 Np-239 2.400E-01 0.260 0.173 0.433 0.1039 Na-24 2.OOOE-03 0.554 4.121 4.675 0.0094 P-32 2.OOOE-05 0.695 0.000 0.695 0.0000 Cr-51 5.OOOE-04 0.004 0.033 0.036 0.0000 Mn-54 4.OOOE-05 0.004 0.836 0.840 0.0000 Mn-56 5.000E-02 0.830 1.692 2.522 0.1261 Co-58 5.OOOE-03 0.034 0.976 1.009 0.0050 Co-60 5.OOOE-04 0.097 2.504 2.601 0.0013 Fe-59 8.OOOE-05 0.118 1.189 1.307 0.0001 Ni-65 3.OOOE-04 0.632 0.549 1.181. 0,0004 Zn-65 2.OOOE-06 0.007 0.584 0.591 0.0000 Zn-69m 3.000E-05 0.022 0.417 0.439 0.0000 Ag-lr0m 6.OOOE-05 0.072 2.751 2.823 0.0002 W-187 3.OOOE-03 0.312 0.481 0.793 0.0024 F-18 4.OOOE-03 0.250 1.022 1.272 0.0051 Total 1.458E+00 Total 1.856E+00 Ai From Reference 10.68, Tables III & IV for all isotopes except noble gases Ai From Table 12 for noble gases Bi & Ci From Reference 10.8, Appendix A for isotope having half life > 15 minutes E-BAR = SUM (Weighted E-Bar)/Sum (Ai) 1.273 100/E-BAR Coolant Concentration . 78.556 Percent Fuel Defect Based on E-BAR = (100/E-BAR)/Sum (Ai) 53.878 CC-AA-309-1001, Rev 3 ]

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO.3 PAGE NO. 71 of 109 Table 14 Noble Gas Concentration Based on 100lE-BAR Normal Noble Gas Noble Gas Noble Gas Scaling Concentration Isotope Activity Factor Based On Concentration Based On 100/E-BAR (iCi/g) 100/E-BAR (jiCi/g)

A B C=AxB Kr-83m 1.518E-03 5.388E+01 8.178E-02 Kr-85m 2.723E-03 5.388E+01 1.467E-01 Kr-85 8,929E-06 5.388E+01 4.810E-04 Kr-87 8.929E-03 5.388E+01 4.810E-01 Kr-88 8,929E-03 5.388E+01 4.810E-01 Xe-131m 6.696E-06 5.388E+01 3.608E-04 Xe-133m 1.295E-04 5.388E+01 6.975E-03 Xe-133 3.661E-03 5.388E+01 1.972E-01 Xe-135m 1.161E-02 5.388E+01 6.254E-01 Xe-135 9.821E-03 5.388E+01 "5.292E-01 Xe-138 3.973E-02 5.388E+01 2.141E+00 A Fro m Table 12 B From Table 13 Table 15 Reactor Coolant Specific Activity Concentration Used In RADTRAD Nuclide Inventory File RCS def.txt Iodine & Reactor Post-LOCA Noble Gas Coolant Activity Isotope Activity Mass Release Concentration ACi/g (g), (Ci)

A B C=AxB/1E6 1-131 4.615E-02 3.708E+08 .1711E+02 1-132 4.260E-01 3.708E+08 .1580E+03 1-133 3.160E-01 3,708E+08 .1172E+03 1-134 8.520E-01 3.708E+08 .3159E+03 1-135 4.615E-01 3.708E+08 .1711E+03 Kr-83m 8.178E-02 3.708E+08 .3032E+02 Kr-85m 1.467E-01 3.708E+08 .5440E+02 Kr-85 4.810E-04 3.708E+08 .1784E+00 Kr-87 4.810E-01 3.708E+08 .1784E+03 Kr-88 4.810E-01 3,708E+08 .1784E+03 Xe-131m 3.608E-04 3.708E+08 .1338E+00 Xe-133m 6.975E-03 3.708E+08 .2586E+01 Xe-133 1.972E-01 3.708E+08 .7313E+02 Xe-135m 6.254E-01 3.708E+08 .2319E+03 Xe-135 5.292E-101 3.708E+08 .1962E+03 Xe-138 2.141E+00 3.708E+08 .7938E+03 A - Iodine Activity Concentration From Table 11 A - Noble Gas Activity Concentration From Table 14 B From Section 7.13 CC-AA-309-1001, Rev 3

CALCULATION NO, H1-ZZ-ADC-1880 REVISION NO.'3 PAGE NO.72 of 109 Table 16 Post-LOCA Cont. Leakage Iodine Activity Deposited on CR Charcoal Filter Iodine Fraction Elemental & 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 3.032E+17 2.301E+17 6.206E+00 1-132 4.420E+14 l.208E-02 3.664E+15 8.289E+00 1-133 3.997E+15 1.662E-01 5.040E+16 1.261E+01 1-134 1.685E+14 5.276E-03 *1.600E+15 9.496E+00 1-135 1.270E+15 5.753E-02 1.745E+16 1.373E+01 Total Iodine Atoms/Activity 3.032E+17 5.033E+01 CC-AA-309-1001, Rev 3

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO.3 PAGENO. 73 of 109 Table 17 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 = BiZB Co-58 1.299E-08 4.084E-16 3.145E-08 4.349E-07 Co-60 1.556E-08 1.376E-14 8.846E-07 1.465E-05 Rb-86 7.099E-07 8.725E-15 1.229E-08 9.290E-06 Sr-89 1.827E-05 6.290E-13 3.442E-08 6.697E-04 Sr-90 1.795E-06 1.316E-I1 7.331E-06 1.401E-02 Sr-91 3.115E-05 8.593E-15 2.759E-10 9.149E-06 Sr-92 1.472E-05 1.171E-15 7.956E-11 1.247E-06 Y-90 3.920E-08 7.205E-17 1.838E-09 7.671E-08 Y-91 2.409E-07 9.821E-15 4.078E-08 1.046E-05 Y-92 3.364E-06 3.496E-16 1.039E-10 3.722E-07 Y-93 2.466E-07 7.392E-17 2.997E-10 7.870E-08 Zr-95 3.295E-07 1.534E-14 4.655E-08 1.633E-05 Zr-97 9.197E-07 4.811E-16 5.231E-10 5.122E-07 Nb-95 3.312E-07 8.469E-15 2.557E-08 9.018E-06 Mo-99 4.245E-06 8.851E-15 2.085E-09 9.424E-06 Tc-99m 3.792E-06 7.212E-16 1.902E-10 7.678E-07

Ru-103 6.536E-06 .2.025E-13 3.098E-08 2.156E-04 Ru-105 1.680E-06 2.499E-16 1.488E-10 2.660E-07 Ru-106 2.499E-06 7.469E-13 2.989E-07 7.952E-04 Rh-105 2.144E-06 2.540E-15 1.185E-09 2.704E-06 Sb-127 4.689E-06 1.756E-14 3.745E-09 1.870E-05 Sb-129 1.047E-05 1.862E-15 1.778E-10 1.983E-06 Te-127 4.696E-06 1.780E-15 3.789E-10 1.895E-06 Te-127m 6.309E-07 " 6.688E-14 1.060E-07 7.121E-05 Te-129 1.190E-05 5.680E-16 4.77SE-11 6.048E-07 Te-129m 2.109E-06 7.001E-14 3.319E-08 7.455E-05 CC-AA-309-1001, Rev 3

I CALCULATION NO. H-I-ZZ-MDC-1880 I REVISION NO.3 PAGE NO. 74 of 109 Table 17 (Cont'd)

Relationship of Aerosol Mass and Activity CR Region P 2.0 hr Aerosol Isotopic Isotope Activity Mass Mass Per Ci Aerosol (Curie) (kg) (kg/Ci) Fraction A Bi Ci = Bi fAi Di = Bi/EB Te-131m 4.487E-05 5.627E-14 1.254E-09 5.991E-05 Te-132 6.364E-05 2.096E-13 3.294E-09 2.232E-04 Cs-134 5.999E-05 4.637E-1 1 7.729E-07 4.937E-02 Cs-136 2.077E-05 2.834E-13 1.364E-08 3.017E-04 Cs-137 7.581E-05 8.716E-10 1.150E-05 9.280E-01 Ba-139 1.231E-05 7.527E-16 6.114E-I1 8.015E-07 Ba-140 3.236E-05 4.420E-13 1.366E-08 4.706E-04 La-140 9.197E-07 1.655E-15 1.799E-09 1.762E-06 La-141 2.155E-07 3.811E-17 1.768E-10 4.058E-08 La-142 1.209E-07 8.447E-18 6.986E-11 8.994E-09 Ce-141 7.713E-07 2.707E-14 3.510E-08 2.882E-05 Ce-143 6.879E-07 1.036E-15 1.506E-09 1.103E-06 Ce-144 1.261E-06 3.954E-13 3.135E-07 4.21GE-04 Pr-143 2.784E-07 4.134E-15 1.485E-08 4.402E-06 Nd-147 1.224E-07 1.514E-15 1.236E-08 1.611E-06 Np-239 .8.660E-06 3.733E-14 4.310E-09 3.974E-05 Pu-238 1.537E-09 8.978E-14 5.841E-05 9.559E-05 Pu-239 1.854E-10 2.982E-12 1.609E-02 3.175E-03 Pu-240 2.397E-10 .1.052E-12 4.388E-03 1.120E-03 Pu-241 6.953E-08 6.750E-13 9.707E-06 7.187E-04 Am-241 3.130E-11 9.120E-15 2.914E-04 9.710E-06 Cm-242 7.410E-09 2.236E-15 3.017E-07 2.380E-06 Cm-244 3.563E-1O 4.405E-15 1.236E-05 4.690E-06 Total 9.392E-10 1.00E+60 A & B From. RADTRAD Run HEPU350CLO2 output file @ 2.0 hr from Control Room Compartment Nuclide Inventory I CC-AA-309-1001, Rev 3

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO. 3 PAGE NO. 75 of 109 Table 18 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 4.349E-07 4.970E-08 2.161E-14 6.873E-07 Co-60 8.846E-07 1.465E-05 4.970E-08 7.282E-13 8.232E-07 Rb-86 1.229E-08 9.290E-06 4.970E-08 4.617E-13 3.757E-05 Sr-89 3.442E-08 6.697E-04 4.970E-08 3.328E-11 9.669E-04 Sr-90 7.331E-06 1.401E-02 4.970E-08 6.965E-10 9.501E-05 Sr-91 2.759E-10 9.149E-06 4.970E-08 4.547E-13 1.648E-03 Sr-92 7.956E-1 1 1.247E-06 4.970E-08 6.197E-14 7.789E-04 Y-90 1..838E-09 7.671E-08 4.970E-08 3.813E-15 2.074E-06 Y-91 4.078E-08 1.046E-05 4.970E-08 5.197E-13 1.275E-05 Y-92 1.039E-10 3.722E-07 4.970E-08 1.850E-14 1.780E-04 Y-93 2.997E-10 7.870E-08 4.970E-08 3.912E-15 1.305E-05 Zr-95 4.655E-08 1.633E-05 4.970E-08 8.117E-13 1.744E-05 Zr-97 5.231E-10 5.122E-07 4.970E-08 2.546E-14 4.867E-05 Nb-95 2.557E-08 9.018E-06 4.970E-08 4.482E-13 1.753E-05 Mo-99 2.085E-09 9.424E-06 4.970B-08 4.684E-13 2.247E-04 Tc-99m 1.902E-10 7.678E-07 4.970E-08 3.816E-14 2.007E-04 Ru-103 3.098E-08 2.156E-04 4.970E-08 1.072E-11 3.459E-04 Ru-105 1.488E-10 2.660E-07 4.970E-08 1.322E-14 8.888E-05 Ru-106 2.989E-07 7.952E-04 4.970E-08 3:953E-11 1.322E-04 Rh-105 1.185E-09 2.704E-06 4.970E-08 1.344E-13 1.134E-04 Sb-127 i. 3.745E-09 1.870E-05 4.970E-08 9.293E-13 2.482E-04 Sb-129 1.778E-10 1.983E-06 4,970E-08 9.855E-14 5.542E-04 Te-127 3.789B-10 1.895E-06 4.970E-08 9.417E-14 2.485E-04 Te-127m 1.060E-07 7.121E-05 4.970E-08 3.539E-12 3..338E-05 Te-129 4.775E-11 6.048E-07 4.970E-08 3.006E-14 6.295E-t4 Te-129m 3.319E-08 7.455E-05 4.970E-08 3.705E-12 1.11 6E-04 CC-AA-309-1001, Rev 3 1

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO.3 PAGE NO.76 of 109 Table 18 (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 5.991E-05 4.970E-08 2.978E-12 2.374E-03 Te-132 3.294E-09 2.232E-04 4.970E-08 1.109E-11 3.368E-03 Cs-134 7.729E-07 4.937E-02 4.970E-08 2.454E-09 3.175E-03 Cs-136 1.364E-08 3.017E-04 4.970E-08 1.499E-11 1.099E-03 Cs-137 1.150E-05 9.280E-01 4.970E-08 4.612E-08 4.012E-03 Ba-139 6.114E- 11 8.015E-07 4.970E-08 3.983E-14 6.515E-04 Ba-140 1.366E-08 4.706E-04 4.970E-08 2.339E-11 1.712E-03 La-140 1.799E-09 1.762E-06 4.970E-08 8.756E-14 4.867E-05 La-141 1..768E-10 4.058E-08 4.970E-08 2,017E-15 1,141E-05 La-142 6.986E-11 8,994E-09 4.970E-08 4;470E-16 6.399E-06 Ce-141 3.510E-08 2.882E-05 4.970E-08 1,432E-12 4,082E-05 Ce-143 1.506E-09 1.1 03E-06 4.970E-08 5.482E-,14 3,640E-05 Ce-144 3.135E-07 4.210E-04 4.970E-08 2093E-I1 6.674E-05 Pr- 143 1.485E-08 4.402E-06 4.970E-08 2.188E-13 1.473E-05 Nd-147 1.236E-08 1.611E-06 4.970E-08 8.009E-14 6.479E-06 Np-239 4.310E-09 3.974E-05 4.970E-08 1.975E-12 4.583E-04 Pu-238 5.841E-05. 9.559E-05 4,970E-08 4.751E-12 8.133E-08 Pu-239 1.609E-02 3.175E-03 4.970E-08 1.578E-10 9.81OE-09 Pu-240 4.388E-03 1.120E-03 4.970E-08 5.567E-11 1.269E-08 Pu-241 9.707E-06 7.187E-04 4.970E-08 3.572E-I1 3 .680E-06 Am-241 2.914E-04 9.710E-06 4.970E-08 4.826E-13 1.656E-09 Cm-242 3.017E-07 2.380E-06 4.970E-08 1,183E-13 3.921E-07 Cm-244 1.236E-05 4.690E-06 4.970E-08 2.331E-13 1.886E-08 Total CR Aerosol Mass/ActiVity @ 720 hrs 4.970E-08 2.383E-02 A & B From Table 17 C From Section 7.6.3A I CC-AA-309-1001, Rev 3 I

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO. 3 ' ] PAGE NO. 77 of 109 Table 19 720-hrs Post-LOCA Total Iodine & Aerosol Activity On CREF Charcoal/HEPA Filters (Ci)

Total Total Isotope Activity Isotope Activity Ci Ci A A Co-58 6.873E-07 Te-131m 2.374E-03 Co-60 8.232E-07 Te-132 3,368E-03 Rb-86 3.757E-05 Cs-134 3.175E-03 Sr-89 9.669E-04 Cs-136 1.099E-03 Sr-90 9.501E-05 Cs-137 4.012E-03 Sr-91 1.648E-03 Ba-139 6.515E-04 Sr-92 7.789E-04 Ba-140 1.712E-03 Y-90 2.074E-06 La-140 4.867E-05 Y-91 1.275E-05 La-141 1.141E-05 Y-92 1.780E-04 La-A42 6.399E-06 Y-93 1.305E-05 Ce-141 4.082E-05 Zr-95 1.744E-05 Ce-143 3.640E-05 Zr-97 4.867E-05 Ce-144 6.674E-05 Nb-95 1.753E-05 Pr-143 1.473E-05 Mo-99 2.247E-04 Nd-147 6.479E-06 Tc-99m 2.007E-04 Np-239 4.583E-04 Ru-103 3.459E-04 Pu-238 8.133E-08 Ru-105 8.888E-05 Pu-239 9.8 IOE-09 Ru-106 1.322E-04 Pu-240 1.269E-08 Rh-105 1.134E-04 Pu-241 3.680E-06 Sb-127 2.482E-04 Am-241 1.656E-09 Sb-129 5.542E-04 Cm-242 -3.921E-07 Te-127 2.485E-04 Cm-244 1.886E-08 Te-127m 3.338E-05 1-131 6.206E+00 Te-129 6.295E-04 1-132 8.289E+0O Te-129m 1.1!6E-04 1-133 1.261E+01 1-134 9.496E+00 1-135 1.373E+01 A - Aerosol From Table 18 A - Iodine From Table 16 I CC-AA-309-1001, Rev 3

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO.3 PAGE NO. 78 of 109 Table 20 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 0.OOOE+00 5.260E-07 8.686E-06 Co-60 9.771E-06 0.OOOE+00 6.298E-07 1.040E-05 Kr-85 1.819E+00 O.OOOE+00 3.369E-01 2.156E+00 Kr-85m 3.592E+01 0.000E+00 6.564E+00 4.248E+01

Kr-87 5.160E+01 *0.OOOE+00 9.074E+00 6.067E+0I Kr-88 9.123E+01 0.OOOE+00 1.654E+01 1.078E+02 Rb-86 2.919E-02 0.OOOE+00 1.677E-04 2.936E-02 Sr-89 1.148E-02 0.OOOE+00 7.402E-04 1.222E-02

Sr-90 " 1.128E-03 0.000E+00 7.268E-05 1.200E-03 Sr-91 2.118E-02 0.OOE+00 1.362E-03 2.254E-02 Sr-92 1.221E-02 O.OOOE+00 7.814E-04 1.299E-02 Y-90 1.597E-05 0.OOOE+00 1.070E-06 1.704E-05 Y-91 1.484E-04 O.OOOE+00 9.577E-06 1.579E-04 Y-92 9.437E-04 0.OOOE+00 6.861E-05 1.012E-03 Y-93 1.669E-04 0.OOOE+00 1.074E-05 1.776E-04 Zr-95 2.071E-04 0.000E+00 1.335E-05 2.204E-04 Zr-97 6.039E-04 0.OOOE+00 3.889E-05 6.428E-04 Nb-95 2.080E-04 0.OOOE+00 1.341E-05 2.214E-04 Mo-99 2.697E-03 0.OOOE+00 1.738E-04 2.871E-03 Tc-99m 2.383E-03 0.OOOE+00 1.536E-04 2.537E-03 Ru-103 4.108E-03 0.OOOE+00 2.648E-04 4.373E-03 Ru-105 1.250E-03 0.OOOE+00 8.023E-05 1.330E-03 Ru-106 1.570E-03 0.000E+00 1.012E-04 1.671E-03 Rh-105 1.351E-03 0.OOOE+00 8.706E-05 1.438E-03 Sbr127 2.970E-03 0.O0OE+00 1.914E-04 3.161E-03 Sb-129 7.829E-03 O.OOOE+00 5.025E-04 8.331E-03 Te-127 2.960E-03 0.OOOE+00 1.908E-04 3.151E-03 Te-127m 3.962E-04 O.OOOE+00 2.554E-05 4.217E-04 Te- 129 8.327E-03 0.000E+00 5.358E-04 8.862E-03

.Te-129m 1.325E-03 O.OOOE+00 8.537E-05 1.410E-03 I CC-AA-309-1001, Rev 3

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO. 3 PAGE NO. 79 of 109 Table 20 (Cont'd) 0-1 Hrý 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 0.OOOE+00 1.862E-03 3.076E-02 Te-132 4.036E-02 0.OOOE+00 2.601E-03 4.296E-02 1-131 1.245E+01 3.802E+00 2.470E-01 1.650E+01 1-132 1.741E+01 5.064E+00 2.965E-01 2.277E+0I 1-133 2.559E+01 7.752E+00 4.969E-01 3.384E+01 1-134 2.547E+01 6.562E+00 2.940E-01 3.232E+01 1-135 2.860E+01 8.511E+00 5.287E-01 3.764E+01 Xe-133 2.918E+02 8.516E-02 5.401E+01 3.459E+02 Xe- 135 1.061E+02 1.104E+00 1.945E+01 1.266E+02 Cs-134 2.460E+00 0.OOOE+00 1.415E-02 2.474E+00 Cs-136 8.550E-01 0.OOOE+00, 4.910E-03 8.599E-01 Cs-137 3.108E+00 0.OOOE+00 1.788E-02 3.126E+00 Ba-139 1.336E-02 0.OOOE+00 8.496E-04 1.421E-02 Ba-140 2.038E-02 0.OOOE+00 1.313E-03 2.169E-02 La-140 3.309E-04 0.000E+00 2.248E-05 3.534E-04 La-141 1.640E-04 0.OOOE+00 1.052E-05 1.745E-04 La-142 1.238E-04. 0.OOOE+00 7.885E-06 1.317E-04 Ce-141 4.847E-04 0.OOOE+00 3.124E-05 5.159E-04 Ce-143 4.421E-04 0.OOOE+00 2.848E-05 4.705E-04 Ce-144 7.923E-04 0.OOOE+00 5.107E-05 8.433E-04 Pr-143 1.745E-04 0.OOOE+00 1.125E-05 1.857E-04 Nd-147 7.713E-05 0.000E+00 4.971E-06 8.210E-05 Np-239 5.512E-03 0.OOOE+00 3.552E-04 5.867E-03 Pu-238 9.653E-07 0.OOOE+00 6.222E-08 1.028E-06 Pu-239 1.164E-07 0.OOOE+00 7.503E-09 1.239E-07 Pu-240 1.506E-07 0.OOOE+00 9.705E-09 1.603E-07 Pu-241 4.367E-05 0.OOOE+00 2.815E-06 4.649E-05 Am-241 1,965E-08 0.OOOE+00 1.267E-09 2.092E-08 Cm-242 4.655E-06 0.000E+00 3.OOOE-07 4.955E-06 Cm-244 2.238E-07 0.OOOE+00 1.443E-08 2.382E-07 A From HEPU350CLO2.oO B From HEPU350ES02.o0 A From HEPU35OMS02.oO I CC-AA-309-1001, Rev 3

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO.3 PAGE NO. 80 of 109 Table 21 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 3.186E-05 1.793E-04 Co-60 1.766E-04 0.OOOE+00 3.816E-05 2.147E-04 Kr-85 2.517E+01 0.000E+00 8.923E+00 34409E+01 Kr-85m 4.369E+02 O.OOOE+00 1.532E+02 5.901E+02 Kr-87 4.569E+02 O.OOOE+00 1.549E+02 6.118E+02 Kr-88 1.031E+03 O.OOOE+00 3.589E+02 1.390E+03 Rb-86 3.603E-02 O.OOOE+00 2.254E-03 3.828E-02 Sr-89 2.075E-01 O.OOOE+00 4.483E-02 2.523E-01 Sr-90 2.038E-02 O.OOOE+00 4.404E-03 2.478E-02 Sr-91 3.635E-01 0.000E+00 7.797E-02 4.415E-01 Sr-92 1.846E-01 0.OOOE+00 3.883E-02 2.234E-01 Y-90 3.812E-04 O.OOOE+00 9.152E-05 4.727E-04 Y-91 2.713E-03 O.OOOE+00 5.894E-04 3.302E-03 Y-92 3.015E-02 0.OOOE+00 7.934E-03 3.809E-02 Y-93 2.874E-03 0.OOOE+00 6.166E-04 3.490E-03 Zr-95 3.742E-03 O.OOOE+00 8.084E-04 4.550E-03 Zr-97 1.060E-02 0.000E+06 2.282E-03 1.289E-02 Nb-95 3.760E-03 0.OOOE+00 8.124E-04 4.572E-03 Mo-99 4.838E-02 O.OOOE+00 1.044E-02 5.883E-02 Tc-99m 4.306E-02 0.OOOE+00 9.304E-03 5.236E-02 Ru-103 7.422E-02 0.OOOE+00 1.604E-02 9.025E-02 Ru-O05 2.024E-02 0.OOOE+00 4.305E-03 2.455E-02 Ru-106 2.837E-02 O.OOOE+00 6.130E-03 3.450E-02 Rh-105 2.437E-02 0.OOOE+00 5.264E-03 2.963E-02 Sb-127 5.339E-02 O.OOOE+00 1.153E-02 6.491-E-02 Sb-129 1.264E-01 O.OOOE+00 2.687E-02 1.533E-01 Te-127 5.338E-02 O.OOOE+00 1.153E-02 6.491E-02 Te-127m 7.162E-03 O.OOOE+00 1.547E-03 8.709E703 Te-129 1.405E-01 0.OOOE+00 3.006E-02 1.706E-0 I Te-129m 2.394E-02 O.OOOE+00 5.174E-03 2.912E-02 CC-AA-309-1001, Rev 3

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO. 3 PAGE NO. 81 of 109 Table 21 (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+7+C Te-131m .5,138E-01 O.OOOE+00 1.108E-01 6.246E-01 Te-132 7.248E-01 0.OOOE+00 1.565E-01 8.813E-01 1-131 1.687E+01 I.674E{01 2.972E+00 3.659E+01 1-132 2.248E+01 1.879E+01 3.194E+00 4.446E+01 1-133 .3.426E+01 3.313E+01 5.838E+00 7.323E+01 1-134 2.830E+01 1.467E+01 1.979E+00 4.495E+01 1-135 3.73 lE+-OI 3.396E+01 5.878E+00 7.714E+01 Xe-133 4.027E+03 1.443E+00 1.427E+03 5.456E+03 Xe-135 1.487E+03 1.763E+01 5.188E+02 2.023E+03 Cs-134 3.037E+00 0,OOOE+00 1.904E-01 3,228E+00 Cs-136 1.055E+00 0.000E+00 6:594E-02 1.121E+00 Cs-137 3,838E+00 0.OOOE+00 2.405E-01 4.079E+00 Ba-139 1.71 OE-0 1 0.OOOE+00 3.499E-02 2.060E-01 Ba-140 3.677E-01 0.OOOE+00 7.943E-02 4.471E-01 La-140 8.622E-03 0.OOOE+00 2.124E-03 1.075E-02 La-141 2.619E-03 0.OOOE+00 5.556E-04 3.174E-03 La- 142 1.642E-03' O.OOOE+00 3.381E-04 1.980E-03 Ce-141 8.758E-03 0.OOOE+00 1.892E-03 1.065E-702 Ce-143 7.872E-03 0.OOOE+00 1.697E-03 9.569E-03 Ce-144 1.432E-02 0.OOOE+00 3.094E-03 1.741E-02 Pr-143 3.158E-03 0.0OOE+00 6.827E-04 3.840E-03 Nd-147 1.391E-03 0.OOOE+00 3.006E-04 1.692E-03 Np-239 9.876E-02 0.000E+00 2.131E-02 1.201E-01 Pu-238 1.745E-05 0.OOOE+00 3.770E-06 2.122E-05 Pu-239 2.104E-06 0.000E+00 4.547E-07 2.559E-06 Pu-240 2.721E-06 0.000E+00 5.880E-07 3.309E-06 Pu-241 7.894E-04 O.OOOE+00 1.706E-04 9.599E-04 Am-241 3.553E-07 O.OOOE+00 7.677E-08 4.321E-07 Cm-242 8,413E-05 0.OOOE+00 1.81E-05 1.023E-04 Cm-244 4.045E-06 0.000E+00 8.741E-07 4.919E-06 A From HEPU350CLO2.oO B From HEPU350ES02.oO A From HEPU350MS02.oO I -CC-AA-309-1001, Rev 3 1

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO. 3 PAGE NO. 82 of 109 Table 22 0-4 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-IC Co-58 3.773E-04 0 .000E+00 3.514E-04 7.287E-04 Co-60 4.520E-04 0.OOOE+00 4.21 1E-04 8.73 1E-04 Kr-85 1.328E+02 0.OOOE+00 1.051E+02 2.380E+02 Kr-85m 1.904E+03 0.OOOE+00 1.456E+03 3.360E+03 Kr-87 1.307E+03 0.000E+00 8.959E+02 2.203E+03 Kr-88 4.049E+03 0.OOOE+00 3.027E+03 7.076E+03 Rb-86 4.531E-02 0.OOOE+00 1.722E-02 6.253E-02 Sr-89 5.309E-01 0.OOOE+00 4.943E-01 1.025E+00 Sr-90 5.217E-02 0.OOOE+00 4.860E-02 1.008E-01 Sr-91 8.823E-01 0.OOOE+00 7.819E-01 1.664E+00 Sr-92 3.962E-01 0.OOOE+00 3.092E-01 7.055E-01 Y-90 1.287E-03 0.OOOE+00 1.577E-03 2.864E-03 Y-91 7.043E-03 0.OOOE+00 6.683E-03 1.373E-02 Y-92 1.047E-01 0.OOOE+00 1.322E-01 2'369E-01 Y-93 6.996E-03 0.OOOE+00 6.218E-03 1.321E-02 Zr-95 9.574E-03 O.OOOE+00 8.916E-03 1.849E-02 Zr-97 .2.633E-02 0.OOOE+00 2.386E-02 5.019E-02 Nb-95 9.624E-03 0.OOOE+00 8.965E-03 1.859E-02 Mo-99 1.229E-01 0.OOOE+00 1.137E-01 2.366E-01 Tc-99m 1.101E-01 0.OOOE+00 1.025E-01 2.126E-01 Ru-103 1.899E-01 0.OOOE+00 1.768E-01 3.667E-01 Ru-105 4.638E-02 0.OOOE+00 3.881E-02 8.519E-02 Ru-106 7.261E-02 0.OOOE+00 6.764E-02 1.402E-01 Rh-105 6.215E-02 0.OOOE+00 5.768E-02 1.19gE-01 Sb-127 1.359E-01 0.OOOE+00 1.260E701 2.619E-01 Sb-129 2.888E-01 0.OOOE+00 2,409E-01 5.297E-01 Te-127 1.363E-01 0.OOOE+00 1.267E-01 2.629E-01 Te-127m 1.833E-02 -0.000E+00 1.708E-02 3.541E-02 Te-129 3.318E-01 0.000E+00 2.858E-01 6.176E-01 Te-129m 6.129E-02 0.OOOE+00 5.709E-02 1.184E-01 CC-AA-309-1001, Rev 3

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO.3 PAGE NO.83 of 109 Table 22 (Cont'd) 0-4 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 1.293E+00 0.OOOE+00 1.186E+00 2.479E+00 Te-132 1.843E+00 0.OOOE+00 1.707E+00 3.550E+00 1-131 2.386E+01 6.742E+01 2.106E+0I 1.123E+02 1-132 2.907E+01 5.688E+01 1.822E+01 1.042E+02 1-133 4.743E+01 1.280E+02 3.971E+0i 2.152E+02 1-134 3.005E+01 2.503E+01 5.725E+00 6.080E+01 1-135 4.940E+01 1.194E+02 3.642E+01 2.052E+02 Xe-133 2.115E+04 1.294E+01 1.671E+04 3.787E+04 Xe-135 7.566E+03 1.425E+02 5.868E+03 1.358E+04 Cs-134 3.823E+00 0.OOOE+00 1.457E+00 5.280E+00 Cs-136 1.326E+00 0.OOOE+00 5.034E-01 1.830E+00 Cs-137 4.830E+00 0.OOOE+00 1.842E+00 6.672E+00 Ba-139 3.184E-01 0.OOOE+00 2.087E-01 5.272E-01 Ba-140 9.396E-01 0.OOOE+00 8.739E-01 1.813E+00 La-140 3.089E-02 0.OOOE+00 3.949E-02 7.038E-02 La-141 5.918E-03 0.OOOE+00 4.883E-03 1.080E-02 La-142 3.146E-03

0.000E+00 2.141E-03 5.287E-03 Ce-141 2.241E-02 0.OOOE+00 2.086E-02 4.327E-02 Ce-143 1.984E-02 0.OOOE+00 1.822E-02 3.806-E-02 Ce-144 3.665E-02 O.OOOE+00 3.414E-02 7.079E-02 Pr-143 8.097E-03 0.OOOE+00 7.560E-03 1.566E-02 Nd-147 3.555E-03 0.OOOE+00 3.305E-03 6.860E-03 Np-239 2.505E-01 0.OOOE+00 2.314E-01 4.819E-01 Pu-238 4.466E-05 O.OOOE+00 4.160E-05 8.627E-05 Pu-239 5.387E-06 0.OOOE+00 5.019E-06 1.041E-05 Pu-240 6.966E-06 0.OOOE+00 6.489E-06 1.346E-05 Pu-241 2.021E-03 0.000E+00 1.882E-03 3.903E-03 Am-241 9.097E-07 0.000E+00 8.476E-07 1.757E-06 Cm-242 2.153E-04 0.OO0E+00 2.005E-04 4.159E-04 Cm-244 1.036E-05 0.OOOE+00

9.646E-06 2.OOOE-05 A From HEPU350CLO2.oO B From HEPU350ES02.oO A From HEPU350MS02.oO I CC-AA-309-1001, Rev 3

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO.3 PAGE NO. 84 of 109 Table 23 0-8 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 4.278E-04 0.OOOE+00 1.098E-03 1.526E-03 Co-60 5.126E-04 0.000E+00 1.317E-03 1.829E-03 Kr-85 5.644E+02 0.OOOE+00 5.816E+02 1.146E+03 Kr-85m 5.606E+03 0.OOOE+00 5.530E+03 1.114E+04 Kr-87 2.046E+03 0.OOOE+00 1.702E+03 3.748E+03 Kr-88 9.938E+03 0.OOOE+00 9.496E+03 1.943E+04 Rb-86 4.729E-02 0.OOOE+00 5.007E-02 9.736E-02 Sr-89 6.019E-01 0.OOOE+00 1.544E+00 2,146E+00 Sr-90 5.916E-02 O.O00E+O0 1.520E-01 2.11IE-01 Sr-91 9.794E-01 0.OOOE+00 2.13 1E+00 3.I11E+00 Sr-92 4.230E-01 0.OOOE+00 6.313E-01 1.054E+00 Y-90 1.643E-03 0.OOOE+00 7.922E-03 9.565E-03 Y-91 8.040E-03 0.OOOE+00 2.173E-02 2.977E-02 Y-92 1.269E-01 0.000E+00 4,740E-01 6.009E-01 Y-93 7,774E-03 0.OOOE+00 1,708E-02 2.486E-02 Zr-95 1.086E-02 0.OOOE+00 2.786E-02 3.871E-02 Zr-97 2.949E-02 0.OOOE+00 6.895E-02 9.844B-02 Nb-95 1.091E-02 0.OOOE+00 2.803E-02 3.894E-02 Mo-99 1.389E-01 0.OOOE+00 3.482E-01 4.871E-01 Tc-99m 1.248E-01 0.000E+00 3.182E-01 4.430E-01 Ru-103 2.153E-01 0.OOOE+00 5.520E-01 7.673E-01 Ru-105 5.047E-02 0.OOOE+00 9.187E-02 1.423E-01 Ru-106 8.233E-02 O.OOOE+00 2.115E-01 2,938E-01 Rh-105 7.030E-02 0.000E+00 1.772E-0 I 2.475E-01 Sb-127 1.537E-01 f,00OE+00 3.881E 5.418E-01 Sb-129 3.140E-01 0.OOOE+00 5.664E-01 8.804E-01 Te-127 1.544E-01 0.OOOE+00 3.930E-01 5.474E-01 Te-127m 2.079E-02 O.OOOE+00 5.341E-02 7.420E-02 Te-129 3.640E-01 0.OOOE+00 7.142E-01 1.078E+00 Te-129m 6.950E-02 O.OOOE+00 1.785E-01 2.479E-01 Rev 33 ~~~1 CC-AA-309-1001, CC-AA-309-1 00 1, Rev I

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO.3 PAGE NO. 85 of 109 Table 23 (Cont'd) 0-8 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 1.456E+00 0.000E+00 3.545E+00 5.OOOE+00 Te-132 2.084E+00 0.000E+00 5.245E+00 7.329E+00 1-131 2.929E+01 2.615E+02 6.930E+01 3.601E+02 1-132 3.197E+01 1.197E+02 4.206E+01 1.938E+-02 1-133 5.679E+01 4.598E+02 1.225E+02 6.391E+02 1-134 3.023E+01 2.976E+0I 7.070E+00 6.705E+01 1-135 5.639E+01 3.602E+02 9.733E+01 5.139E+02 Xe-133 8.866E+04 1.115E+02 9.126E+04 1.800E+05 Xe-135 2.738E+04 9.991E+02 2.779E+04 5.617E+04.

Cs-134 3.991E+00 0.OOOE+00 4.250E+00 8.240E+00 Cs-I36 1.384E+00 0;000E+00 1.462E+00 2.846E+00 Cs- 137 5.043E+00 0.OOOE+00 5.370E+00 1.041E+01 Ba-139 3.296E-01 *0OOOE+00 3.227E-01 6.523E-01 Ba-140 1.065E+00 0.OOOE+00 2.720E+00 3.785E+00 La-140 4.018E-02 0.OOOE+00 2.067E-01 2.469E-01 La-141 6.412E-03 0.OOOE+00 1.120E-02 1.761E-02 La- 142 3.273E-03 0.000E+00 3.481E-03 6.754E-03 Ce-141 2.540E-02 0.OOOE+00 6.515E-02' 9.056E-02 Ce-143 2.235E-02 0.000E+00 5.469E-02 7.704E-02 Ce-144 4.156E-02 0.OOOE+00 1.067E-01 1.483E-01 Pr-143 9.189E-03 0.OOOE+00 2.377E-02 3.296E-02 Nd-147 4.027E-03 0.OOOE+00 1.028E-02 1.431E-02 Np-239 2.830E-01 0.OOOE+00 7.065E-01 9.895E-01 Pu-238 5.064E-05 0.000E+00 1.301E-04 1.807E-04 Pu-239 6.109E-06 0.OOOE+00 1.570E-05 2.181E-05 Pu-240 7.899E-06 0.OOOE+00 2.029E-05 2.819E-05 Pu-241 2.291E-03 0.OOOE+00 5.886E-03 8.177E-03 Am-241 1.032E-06 0.000E+00 2.653E-06 3.684E-06 Cm-242 2.441E-04 0.OOOE+00 6.269E-04 8.710E-04 Cm-244 1.174E-05 0.OOOE+00 3.016E-05 4.191E-05 A From HEPU350CLO2.oO B From EMPU350ES02.o0 A From BEPU350MS02.o0 I CC-AA-309-1001, Rev 3

CALCULATION NO. H-I-ZZ-MDC-1880 REVISION NO. 3 PAGE NO. 86 of 109 Table 24 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 4.309E-04 O.OOOE+00 2.516E 2.947E-03 Co-60 5.163E-04 0.OOOE+00 3.056E-03 3.572E-03 Kr-85 9.035E+04 0.OOOE+00 1.129E+05 2.032E+05 Kr-85m 1.309E+04 O.OOOE+00 1,416E+04 2,725E+04 Kr-87 2.201E+03 0.OOOE+00 1.881E+03 4.082E+03 Kr-88 1.589E+04 0.OOOE+00 1.646E+04 3.235E+04 Rb-86 4.741E-02 0.OOOE+00 1.084E-01 1.558E-01 Sr-89 6.062E-01 0.000E+00 3.522E+00 4.128E+00 Sr-90 5.959E-02 0.OOOE+00 3.529E-01 4.124E-01 Sr-91 9.835E-01 0.OOOE+00 3.307E+00 4.290E+00 Sr-92. 4.235E-01 0.OOOE+00 7.283E-01 1.152E+00 Y-90 1.696E-03 0.OOOE+00 6.526E-02 6.695E-02 Y-91 8.106E-03 0.OOOE+00 5.412E-02 6.223E-02 Y-92 1.280E-01 0.OOOE+00 7.327E-01 8.606E-01 Y-93 7.808E-03 0.OOOE+00 2.684E-02 3,465E-02 Zr-95 1,093E-02 0.OOOE+00 6.376E-02 7.469E-02 Zr-97 2.965E-02 0.OOOE+00 1.188E-0I 1.484E-01 Nb-95. 1.099E-02 O.OOOE+00 6.495E-02 7.594E-02 Mo-99 1.398E-01 0.OOOE+00 6.973E-01 8.371E-01 Tc-99m 1.256E-01 O.OOOE+00 6.542E-0I 7.799E-01 Ru-103 2.168E-01 0.OOOE+00 1.253E+00 1.470E+00 Ru-105 5.058E-02 O.OOOE+00 1.192E-01 1,697E-01 Ru-106 8.293E-02 0.OOOE+00 4.898E-01 5,727E-01 Rh-105 7.076E-02 0.OOOE+00 3.432E-01 4.139E-01 Sb-127 1.548E-01 0.000E+00 7.967E-01 9.514E-0 1 Sb-129 3.147E-01 0.OOOE+00 7.296E-01 1.044E+00 Te-127 1.554E-01 0.OOOE+00 8.315E-01 9.869E-01 Te-127m 2.094E-02 O.OOOE+00 1.236E-01 1,445E-01 Te-129 3.651E01 0.000E+00 1.074E+00 1.439E+00 Te-129m 7.OOOE-02 0.OOOE+00 1 4.044E-01 4.744E-01 CC-AA-309-1001, Rev 3

CALCULATION NO. H-1-ZZ-MDC-1880 I REVISION NO. 3 1 PAGE NO. 87 of 109 Table 24 (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 6.589E+00 8.053E+00 Te-132 2.098E+00 0.OOOE+00 1.064E+01 1.274E+01 1-131 1.79IE+02 2.988E+04 1.832E+03 3.189E+04 1-132 3.351E+01 1.641E+02 9.01IE+01 2.877E+02 1-133 9.793E+01 5.499E+03 5.394E+02 6.136E+03 1-134 3.023E+01 3.010E+01 7.132E+00 6.747E+01 1-135 6.644E+01 1.221E+03 2.031E+02 1.491E+03 Xe-133 3.945E+06 6.750E+04 4.738E+06 8.750E+06 Xe-135 1.260E+05 1.599E+04 1.379E+05 2.799E+05 Cs-134 4.001E+00 0.OOOE+00 9.561E+00 1.356E+01 Cs-136 1.388E+00 0.OOOE+00 3.127E+00 4.514E+00 Cs-137 5.056E+00 -0.000E+00 1.210E+01 1.715E+01 Ba-139 3.297E-01 0.000E+00 3.338E-01 6.635E-01 Ba-140 1.072E+00 0.OOOE+00 5.966E+00 7.038E+00 La-140 4.140E-02 0.OOOE+00 1.263E+00 1.304E+00 La-141 6,425E-03 0.OOOE+00 1.409E-02 2.051E-02 La-142 3.274E-03 0.OOOE+00 3.645E-03 6.920E-03 Ce-141 2.559E-02. 0.OOOE+00 1.474E-01 1.729E-01 Ce-143 2.248E-02 0.OOOE+00 1.027E-01 1.252E-01 Ce-144 4.186E-02 0.000E+00 2.470E-01 2.88SE-01 Pr-143 9.257E-03 0.OOOE+00 5.422E-02 6.348E-02 Nd-147 4.056E-03 0.OOOE+00 2.240E-02 2.646E-02 Np-239 2.848E-01 0.OOOE+00 1.397E+00 1.682E+00 Pu-238 5.101E-05 0,000E+00 3.021E-04 3.532E-04 Pu-239 .. ,153-E-06 0.600E+00 3.652E-05 4.267E-05 Pu-240 7.956E-06 0.OOOE+00 4.712E-05 5.508E-05 Pu-241 2.308E-03 0.OOOE+00 1.36713-02 1.597E-02 Am-241 1.039E-06 0.OOOE+00 6.246E-06 7.285E-06 Cm-242 2.459E-04 0.000E+00 1.447E-03 1.693E-03 Cm-244 1.183E-05 0.OOOE+00 7.003E-05 8.186E-05 A From HEPU350CLO2.oO B From HEPU350ES02.o0 A From HEPU350MS02.o0 I CC-AA-309-1001, Rev 3

ICALCULATION NO. H--ZZ-MDC-1880 IREVISION NO.3 PAGE NO.88 of 109 1 Table 25 Containment Isolation Valve Expected To Remain Open for 120 sec Following A LOCA Existing Proposed Valve Penetration Valve Maximum Maximum Reference Function No. Number Isolation Isolation P&ID Time Time (Seconds) (Seconds)

Main Steam Line Drain. P12 HV-F016 (AB-V039) 30 120. M-41-!

Isolation Valves P12 HV-F019 (AB-V040) 30 120 Sht 1 Reactor Recirculation Water P17 BB-SV-4310 15 120 M-43-1 Sample Line Isolation . Sht 1I P17 BB-SV-4311 15 120 Valves RHR Shutdown Cooling P3 HV-F008 (BC-V164) 45 12 M-51-1 Suction Isolation Valve 0 Sht 1 RHR to Suppression P214B HV-F027A (BC-V1l2) 75 120 M-51-1 Chamber Spray Header Sht 1 & 2 Isolation Valves P214A HV-F027B (BC-V015) 75 120 Sht_1_&_2 RHR Shutdown Cooling P4B HV-FO15A (BCmVI10) 45

120 M-51-1 Return Isolation Valves P4A HV-FO15B (BC-V013) 45 120 Sht 1 & 2 Core Spray Test to P217B HV-F015A(BE-V025) 80 120 Suppression Pool Isolation M-52-1 Valves P217A HV-FO15B (BE-V026) 80 120 RWCU Supply Isolation P9 HV-FOOI (BG-VOOI) 45 120 M-44-1 Valves P9 HV-F004 (BG-VOV02) . 45 120 . Sht 1 TWC Suction Isolation P223 HV-4680 (EE-V003) 45 120 M-53-1 Valves P223 HV-4681 (EE-V004) 45 120 Sht 2 TWC Return Isolation P222 . HV-4652 (EE-V002) 45 120 M-53-1 Valves P222 HV-4679 (EE-VOO1) 45 120 Sht 2.

Drywell Floor Drain Sump P25 HV-F003 (HB-VO05)

30 120 M-61-1 Discharge Isolation Valves P25 HV-F004 (HB-V006) 30 120 Sht 1 Drywell Equipment Drain, P26 HV-F0!9.(I*-VO45) 30 120 M-61-1 Sump Discharge Isolation Sht 2 Valves P26 HV-F020 (HB-V046) 30 120 P8B HV-9531B1 (GB-V081) 60 120 Chilled Water to Drywell P38A HV-953 1B3 (GB-V083) 60 . 120 M-87-1 Coolers Isolation Valves P8B HV-9531AI (GB-V048) 60 120 Sht 2 P38A HV-9531A3 (GB-V070) 60 120 Chilled Water From P8A HV-9531B2 (GB-V082) 60 120 Drywell Coolers Isolation P38B HV-953 IB4 (GB-V084) 60 120 M-87-1 P8A HV-9531A2 (GB-V046) 60 120 Sht 2 Valves P38B HV-9531A4(GB-V071) 60 120 Recirculation Pump Seal P19 HV-3800A (BF-V098) 45 120 M-43-1 Water Isolation Valves P20 HV-3800B (BF-V099) 45 120 Sht 1 I CC-AA-309-1001, Rev 3

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO. 3 PAGE NO. 89 of 109 Table 25 (Cont'd)

Containment Isolation Valve Expected To Remain Open for 120 sec Following A LOCA Existing Proposed Valve Penetration Valve Maximum Maximum Reference Function No. Number Isolation Isolation P&ID Time Time (seconds) (seconds)

Drywell Purge Supply P22 HV-4956 (GS-V009) 5 120 M-57-1 Isolation Valves P22/P220 HV-4979 (GS-V021) 5 120 Sht 1 P23 H-v-4951 (Gs-v025) 15 120 Drywell Purge Exhaust M-57-1 Isolation Valves P23 HV-4950 (GS-V026) 5 120 Sht 1 P23 HV-4952 (GS-V024) 5 120 Suppression Chamber Purge P22/P220 HV-4980 (GS-V020) 5 120 M-57-1 Supply Isolation Valves P220 HV-4958 (GS-V022) 5 120 Sht I Suppression Chamber Purge P219 HV-4963 (GS-V076) 15 120 M-57-1 Exhaust Isolation Valves P219 HV-4962 (GS-V027) *5 120 Sht 1 P219 HV-4964 (GS-V028) 5 120 Nitrogen Purge Isolation J7D/J202 H-V-4974 (GS-V053) 45 120 M-57-1 Valves P221P220 HV-4978 (GS-V023) 5 120 Sht 1 J9E [I{V-4955A (GS-V045) .45 120 BE HV-4955A (GS-V045) 45 120 Drywell -1H2/02 Analyzer Inlet Isolation Valves (Loop J9E HV-4983A (GS-V046) 45 120 M-57-1 A) J10C HV-4984A (GS-V048) 45 120 Sht 1 J10C HV-5019A (GS-V047) 45 120 J3B HV-4955B (GS-V031) 45 120 Drywel' H2/02 Analyzer J3B HV-4983B (GS-V032) 45 120 M-57-1 Inlet Isolation Valves (Loop B) J7D/J202 HV-4984B (GS-V034) 45 120 Sht 1

J7D HV-5019B (GS-V033) 45 120 J212 HV-4965A (GS-V050) 45 120 Suppression Chamber H2/02 Analyzer Inlet J212 HV-4959A (GS-V049) 45 120 M-57-1 Isolation Valves J210 HV-4965B (GS-V041) 45 120 Sht 1 J210 HV-4959B (GS-V040) 45 120 5201 HV-4966A (GS-V051) 45 120 J201 HV-5022A (GS-V052) 45 120 M-57-1 H2/02 Analyzer Return to Suppression Chamber J202 HV-4966B (GS-V042) 45 120 Sht 1 Isolation Valves J202/J7D HV-5022B (GS-V043) 45 120 P23 HV-5050A (GS-V002) 45 120 120 P23 HV-5050A (GS-V03) 45 Containment Hydrogen Recombination (CHR) P23 HV-5052A (GS-V003) 45 120 M-58-1 Supply Isolation Valves P22 HV-5050B (GS-V004). 45 120 P22 HV-5052B (GS-V005), 45 120 P220 HV-5053A (GS-V008) 45 120 Containment Hydrogen P220 HV-5054A (GS-V010) 45 120 Recombination (CHR) M-58-1 Return Isolation Valves P219 HV-5053B (GS-V006) 45 120 P219 HV-5054B (GS-V007) 45 120 I CC-AA-309-1001, Rev 3

CALCULATION NO. I-1-ZZ-MDC-1880 REVISION NO.3 PAGE NO.90 of 109 Table 25 (Cont'd)

Containment Isolation Valve Expected To Remain Open for 120 sec Following A LOCA Existing Proposed Valve Penetration Valve Maximum Maximum Reference Function No. Number Isolation Isolation P&ID Time Time (seconds) (seconds)

P28B HV-5152A (KL-V028) 45 120 PCIGS Drywell Supply P28B HV-5126A (KL-V027) 45 120 M-59-1 Header Isolation Valves P28A HV-5152B (KL-V026) 45 120 Sht I P28A HV-5126B (,KL-V025) 45 120 P39 H-V-5148 (Y-L-V001) 45 120 M5-PCIGS Drywell Suction P39 HV-5147 (KL-V002) 45 120 M-59-1 Isolation Valves Sht I P39 HV-5162 (KL-V049) 45 120 PCIGS Suppression J211 HV-5154 (KL-V018) 15 120 M-59-1 Chamber Supply Isolation Sht 1 Valves J211 HV-5155 (KL-V019) 15 120 Reactor Auxiliaries Cooling P29 HV-2554 (ED-V020) 45 120 System Supply Isolation M-13-1 Valves P29 HV-2553 (ED-V019) 45 120 Reactor Auxiliaries Cooling P30 HV-2556 (-D-V022) 45 120 3 System Return Isolation M-13-1 Valves P30 HV-2555 (ED-V021) 45 1.20 P34A SV-JO04A-1 (SE-V026) 15 120 P34B SV-J004A-2 (SE-VO27) 15 120 Traversing In-Core Probe P34B SV-JOO4A-2 (SE-V027) 15 120 M-59-1 Guide Tube Isolation Valves P34C SV-J004A-3 (SE-V028) 15 120 Sht 3 P34D SV-JOO4A-4 (SE-V029) 15 120 P34E SV-JO04A-5 (SE-V030) 15 120 TIP Purge System Isolation P34F HV-5161 (SE-V004) M-59-1 Valve 15 120 Sht 3 Drywell Leak Detection Radiation Monitoring J8C HV-5018 (SK-VO05) 45 120 M-25-1 System Inlet Isolation Valves , JSC HV-4953 (SK-V006) 45 120 Drywell Leak Detection Radiation Monitoring B5A HV-4957 (SK-VO08) 45 120 M-25-1 System Return Isolation i Valves J5A HV-4981 (SK-V009) 45 120 I CC-AA-309-1001, Rev 3

CALCULATION NO. H-I-ZZ-MDC-1880 REVISION NO. 3 ] PAGE NO. 91 of 109 12.0 FIGURES Figure 1: Containment Leakage RADTRAD Nodalization F -- CC-AA-309-1001, Rev 3

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO. 3 PAGE NO. 92 of 109 Figure 2: HCGS ESF Leakage RADTRAD Nodalization CC-AA-309-1001, Rev 3

I CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO. 3 1 PAGE NO. 93 of 109 Figure 3: HCGS MSIV Leakage Path Volumetric Distribution I . CC-AA-309-1001, Rev 3

I CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO. 3 PAGE NO. 94 of 109**

Figure 4: HCGS MSIV Leakage RADTRAD Nodalization I CC-AA-309-1001, Rev 3 .

I CALCULATION NO. H-I-ZZ-MDC-1880 I REVISION NO. 3 1 PAGE NO.95 of 109 Figure 5- HCGS Control Room RADTRAD Nodalization I CC-AA-309-1001, Rev 3

I CALCULATION NO. H-I-ZZ-MDC-1880 I REVISION NO. 3 PAGE NO.96 of 109 EL 158'-9",

CR Charcoal Filter Bed 3Y (L) x 3' (H) x 4' (W) 7y EL 155'-9".*

CEL 15R'-3" CR .Receptor Location (EL 143'-0")

x.

Figure 6 - CR Filter Shine Dose (Elevation View)

.[ CC-AA-309-1001, Rev 3

I CALCULATION NO. H-1-ZZ-MDC-1880 -REVISION NO. 3 7 PAGE NO. 97 of 109 X Indicates Dose Point Location Figure 7 - CR Filter Shine Dose (Plan View)

I CC-AA-309-1001, Rev 3

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO.3 PAGE NO.98 of 109 FIGURE 8: Relative Locations of FRVS Vent, Purge Vent & CR Air Intake Release Distance To Release Point Point Rece tor Height

_ ft meter ft meter Purge Vent 256.35 78.16 118.00 35.98 FRVS Vent 169.53 51.69 198.67 60.6 I CC-AA-309-1001r, Rev 3

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO. 3 PAGE NO. 99 of 109 Figure 9: Containment Purge Exhaust Release RADTRAD Nodalization CC-AA-309-1001, Rev 3

I CALCULATION NO. H-1-ZZ-MDC- 880 I REVISION NO.3 PAGE NO. 100 of 109 i Settling Velocity Vs Percentile Uncertainty In Velocity Distribution 0.001(

6T

.0.001 1..

0.00112I -

0.00]

c~.

0 0.00088 -

0.0006 6 y~ 0.05 E0 0.0004 0.000:

0 0.1 0.2 0.3 0.4 0.5 0.6 07 Percentile Uncertainty in Velocity Distribution (%)

Figure 10: Settling Velocity Trend With Respect To Percentile Uncertainty In Velocity Distribution I CC-AA-309-1001, Rev 3

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION.NO. 3 PAGE NO. 101 of 109 13.0 AFFECTED DOCUMENTS:

HC Calculation H-1-ZZ-MDC-1880, Rev 2 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 6.2-16'

5. 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. HC Technical Requirements Manual Table 3.6.3-1

11. A list of the I&C Primary Containment Isolation Valve Testing Procedures affected by the increased valve closure time is provided by Hope Creek Valve Program Management, which is included as an affected operating procedure due to the proposed activity. This list is shown in Attachment 14.3 and requires an independent verification from the Hope Creek Mechanical Engineering.

14.0 ATTACHMENTS 14.1 E-mail

Subject:

FRVS Vent Charcoal Filter Efficiencies 14.2 Hope Creek I&C Operating Procedure for Stroke Time Closure (STC) Testing 14.3 I CD with the following electronic files.:

Calculation No: H- 1-ZZ-MDC- 1880, Rev 3.

RADTRAD computer runs CC-AA-309-1001, Rev 3

CALCULATION NO. H-I-ZZ-MDC-1880 REVISION NO. 3 PAGE NO. 102 of 109 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 I 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 methyliodide 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 I CC-AA-309-1001, Rev 3

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO. 3 PAGE NO, 103 of 109 Attachment 14.2 Hope Creek I&C Operating Procedure for Stroke Time Closure (STC) Testing PROCEDURE STEP NO PV COMPONENT EXAM SEQUENCE D S OP-IS.BF-101 1 V IBFHV-FO10 SIC FALSE OP-IS.BF-101 3 V 1BFHV-F180 STC FALSE OP-IS.AB-101 I V 1ABHV-F071 STC FALSE OP-IS.AB-101 2 V 1ABHV-F016 STC FALSE OP-IS.AB-101 3 V IABHV-F019 STC FALSE OP-IS.AB-102 09 V 1ABHV-F022A STC 0 FALSE OP-IS.AB-102 11 V 1ABHV-F022B STC 0 FALSE OP-IS.AB-102 13 V IABHV-F022C STC 0 FALSE OP-IS.AB-102 15 V. 1ABHV-F022D STC 0 FALSE OP-IS.AB-102 17 VI 1ABHV-F028A STC 0 FALSE OP-IS.AB- 102 19 V IABHV-F028B STC 0 FALSE OP-IS.AB-102 21 V iABHV-F028C STC 0 FALSE OP-IS.AB-102 23 V 1ABHV-F028D STC 0 FALSE OP-IS.BB-102 23 V IBBSV-4310 STC FALSE OP-IS.BB-101 3 V 1BBSV-4311 STC FALSE OP-IS.BC-101 I V 1BCSV-F079A STC FALSE OP-IS.BC-101 3 V IBCSV-F078A STC FALSE OP-IS.BC-101 5 V IBCSV-FO07A STC FALSE-OP-IS.BC-101 9 V *IBCHV-FO04A STC FALSE OP-IS.BC-101 15 V 1BCHV-FO03A STC FALSE OP-IS.BC-10I 17 V 1BCHV-F043A SIC FALSE7 OP-1S.BC-101 17 V IBCHV-FO48A STC FALSE OP-IS.BC-10l 23 V 1BCHV-FO16A STC FALSE OP-IS.BC-101 27 V. 1BCHV-FO21A STC FALSE OP-IS.BC-101 33 V 1BCHV-F006A STC FALSE OP-IS.BC-101 37 V 1BCHV-F027A STC FALSE OP-IS.BC-102 41 V 1BCHV-F024A STC FALSE OP-IS.BC-102 01 V IBCSV-FO79B STC 0 FALSE OP-IS.BC-102 03 V 1BCSV-F080B STC 0 FALSE OP-IS.BC-102 05 V 1BCHVF007B STC 0 FALSE OP-IS.BC-102 09 V 1BCHV-FO04B STC 0 FALSE OP-IS.BC-102 15 V 1BCHV-FO03B STC 0 FALSE OP-IS.BC-102 17 V IBCHV-F048B STC 0 FALSE OP-IS.BC-102 23 V 1BCHV-F016B STC 0 FALSE OP-IS.BC-102 27 V 1BCHV-F021B STC 0 FALSE OP-IS.BC-102 31 V 1BCHV-F040 STC 0 FALSE OP-IS.BC-102 33 V 1BCHV-F049B STC 0 FALSE OP-IS.BC-102 37 V 1BCHV-F006B STC 0 FALSE OP-IS.BC-102 41 V 1BCHV-F027B STC 0 FALSE OP-IS.BC-102 45 V 1BCHV-FO24B STC 0 FALSE OP-IS.BC-104 1 V 1BCHV-FO04D STC FALSE OP-IS.BC-104 3 V 1BCHV-FO07D STC FALSE OP-IS.BC-104 7 V 1BCHV-FO0OB STC FALSE OP-IS.BC-105 75 V 1BCHV-F008 STC 0 FALSE OP-IS.BC-105 79 V 1BCHV-F009 STC 0 FALSE OP-IS.BC-105 02A V iBCHIV-F215A STC 0 FALSE OP-IS.BC-105 05 V 1BCHV-F122A STC 0 FALSE OP-IS.BC-105 1 07 V 1BCC-IV-F146A, STC 0 FALSE

[ CC-AA-309-1 00 1, Rev 3

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO.3 PAGE NO. 104 of 109 PROCEDURE STEP NO PV COMPONENT EXAM SEQUENCE D S OP-IS.BC-105 11 V 1BCHV-FO17A STC 0 FALSE OP-IS.BC-105 25 V 1BCHV-FO15B STC 0 FALSE OP-IS.BC-105 29 V IBCHV-F122B STC 0 FALSE OP-IS.BC-105 35 V 1BCHV-F14613 STC 0 FALSE OP-IS.BC-105 39 V 1BCHV-F117B STC 0 FALSE OP-IS.BC-105 52 V 1BCHV-F146C STC 0 FALSE OP-IS.BC-105 56 V IBCHV-FOI7C STC 0 FALSE OP-IS.BC-105 64 V 1BCHV-F146D STC 0 FALSE OP-IS.BC-105 68 V 1BCHV-FO17D STC 0 FALSE OP-IS.BD-101 0] V 1FCHV-F007 STC 0 FALSE OP-IS.BD-101 02 V 1FCHV-F007 STC 0 FALSE OP-IS.BD-101 04 V IFCHV-F076 STC 0 FALSE OP-IS.BD-101 05 V IFCHV-FO25 STC 0 FALSE OP-IS.BD-101 09 V 1FCHV-F084 STC 0 FALSE OP-IS.BD-101 11 V IFCHV-F062 STC 0 FALSE OP-IS.BD-101 18 V 1BDSV-4405 STC 0 FALSE OP-IS.BD-101 21 V 1BDSV-F019 STC 0 FALSE OP-IS.BD-101 23 V 1FCHV-F059 STC 0 FALSE OP-IS.BD-101 29 V 1FCHV-F045 STC 0 FALSE OP-IS.BD-101 33 V 1BDHV-F013 STC 0 FALSE OP-IS.BD-101 37 V 1BDHV-F031 STC 0 FALSE OP-IS.BD-101 39 V 1BDHV-FO10 STC 0 FALSE OP-IS.BD-101 45 V IBDHV-F022 STC 0 FALSE OP-ISBE-101 3. V 1BEHV-F021A STC FALSE OP-IS.BE-101 11 V 1BEHIV-F031A STC 0 FALSE OP-IS.BE-10! 9 V 1BEHV-FO05A STC 0

FALSE OP-IS.BE-101 13 V IBEHV-FO15A STC 0 FALSE OP-IS.BE-102 3 V 1BEHV-F01B STC FALSE OP-IS.BE-102 5 V IBEHV-FOOID STC FALSE OP-IS.BE-102 9 V 1BEHV-FO05B STC 0 FALSE OP-IS.BE-102 11 V IBEHV-FO31B STC 0 FALSE OP-IS.BE-102 13 V 1BEHV-FO15B STC 0 FALSE OP-IS.BE-103 01 V 1BEHV-F039A STC 0 FALSE OP-IS.BE- 103 05 V IBEHV-F039B STC 0 FALSE OP-IS.BF-101 7 *V 1BFHV-F181 STC FALSE OP-IS.BF-102 1 V 1BFHV-3800A STC FALSE OP-IS.BF-102 3 V 1BFHV-3800B STC FALSE OP-IS.BH-10i 1 V IBH-HV-F006A STC FALSE OP-IS.BH-101 .3 V 1BHHV-FO06B STC FALSE OP-IS.BJ-101 01 V 1FDHV-F002 STC 0 FALSE OP-IS.BJ-101 02 V IFDHV-F003 STC 0 FALSE OP-IS.BJ-101 04 V 1FDHV-FI00 STC 0 FALSE OP-IS.BT-101 05 V 1FDHV-F028 STC 0 FALSE OP-IS.BJ- 101 07 V 1FDHV-F029 STC 0 FALSE OP-IS.BJ-101 09 V IFDHV-F075 STC 0 FALSE OP-IS.BJ-101 11 V 1FDHV-F079 STC 0 FALSE OP-IS.BJ-101 13 V 1BJHV-4865 STC 0 FALSE OP-IS.BJ-101 15 V 1BJHV-4866 STC 0 FALSE OP-IS.BJ-101 17 V 1BJHV-4803 STC 0 FALSE OP-IS.BJ-101 19 V IBJHV-4804 STC 0 FALSE CC-AA-309-1001, Rev 3

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO. 3 PAGE NO. 105 of 109 PROCEDURE STEP NO PV COMPONENT EXAM SEQUENCE D S OP-IS.BJ-101 23 V 1APHV-F011 STC 0 FALSE OP-IS.BJ-101 27 V 1BJHV-F012 STC 0 FALSE OP-IS.BJ-101 29 V IFDHV-F071 STC 0 FALSE OP-IS.BJ-101 37 V 1BJHIV-F042 STC 0 FALSE OP-IS.BJ-101 43 V IBJHV-F008 STC 0 FALSE OP-IS.BJ-101 47 V IBJI-V-F006 STC 0 FALSE OP-IS.BJ-101 50 V 1BJIHV-8278 STC 0 FALSE OP-IS.EA-101 01 V 1EAHV-2198C STC 0 FALSE OP-IS.EA-101 05 V IEAHV-2197A STC. 0 FALSE OP-IS.EA-101 09 V 1EAHV-2203 STC 0 FALSE OP-IS.EA-101 39 V 1EASV-2235 STC 0 FALSE OP-IS.EA-101 23 V IEAHV-2198A STC 0 FALSE OP-IS.EA-101 27 V 1EAHV-2197C STC 0 FALSE OP-IS.EA-101 42 V 1EASV-2237 STC 0 FALSE OP-IS.EA-103 01 V 1EAHV-2357A STC 0 FALSE OP-IS.EA-103 03 V IEAHV-2207 STC 0 FALSE OP-IS,EA-103 05 V IEAHV-2346 STC 0 FALSE OP-IS.EA-103 07 V 1EAHV-2357B STC 0 FALSE OP-IS.ED-101 01 V 1EDHV-2553 STC FALSE OP-IS.ED-101 03 V 1EDHV-2554 STC FALSE OP-IS.ED-101 05 V IEDHV-2555 STC FALSE OP-IS.ED-101 07 V 1EDHV-2556 STC FALSE OP-IS.ED-101 09 V 1EDHV-2598 STC FALSE OP-IS.ED-101 11 V IEDHV-2599 STC FALSE

.OP-IS.EE-101 01 V 1EEHV-4679 STC FALSE OP-IS.EE-101 03 V 1EEHV-4652 STC FALSE OP-IS.EE-101 05 V 1EEHV-4680 STC FALSE OP-IS.EE-101 07 V IEEHV-4681 STC FALSE OP-IS.EG-001 02 V 1EGHV-2496A STC TRUE OP-IS.EG-001 04 V 1EGHV-2522A STC TRUE OP-IS.EG-002 02 V IEGHV-2496B STC 0 TRUE OP-IS.EG-002 04 V 1EGHV-2522B STC 0 TRUE OP-IS.EG-003 02 V 1EGHV-2496C STC 0 TRUE OP-IS.EG-003 04 -V 1EGHV-2522C STC 0 TRUE OP-1S.EG-004 02 V 1EGHV-2496D STC 0 TRUE OP-IS.EG-004 04 V IEGHV-2522D STC 0 TRUE OP-IS.EG-101 03 V 1EGHV-2453A STC 0 FALSE OP-IS.EG-101 05 V IEGHV-2321A STC 0 FALSE OP-IS.EG-101 09 V 1EGHV-2452A STC 0 FALSE OP-IS.EG-101 13 V IEGHV-2320A STC 0 FALSE OP-IS.EG-101 15 V 1EGHV-2317A .STC 0 FALSE OP-IS.EG-101 17 V IEGHV-7922A STC 0 FALSE OP-IS.EG-I01 46 V IEGHV-2457A STC 0 FALSE OP-ISEG-I01 49 V 1EGSV-2281-1 STC 0 FALSE OP-IS.EG-102 05 V 1EGHV-2321B STC 0 FALSE OP-IS.EG-102 09 V IEGHV-2452B STC 0 FALSE OP-IS.EG-102 13 V IEGHV-2320B STC 0 FALSE OP-IS.EG-102 15 V IEGHV-2317B. STC 0 FALSE OP-IS.EG-102 17 V 1EGHV-7922B STC 0 FALSE OP-IS.EG-102 46 V 1EGHV-2457B STC 0 FALSE CC-AA-309-1001, Rev 3

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO. 3 PAGE NO. 106 of 109 PROCEDURE STEP NO PV COMPONENT EXAM SEQUENCE D S OP-IS.EG-103 1 V 1EGHV-2522E STC FALSE OP-IS.EG-103 3 V IEGHV-2522F STC FALSE OP-IS.GB-101 1 V IGBHV-9531A-1 STC FALSE OP-IS.GB-101 3 V 1GBHV-9531B-1 STC FALSE OP-IS.GB-101 5 V IGBHV-9531B-2 STC FALSE OP-IS.GB-101 7 V IGBHV-9531A-2 STC FALSE OP-IS.GB-101 9 V 1GBHV-9531A-3 STC FALSE OP-IS.GB-101 11 V 1GBHV-9531B-3 STC FALSE OP-IS.GB-101 13 V 1GBHV-9531A-4 STC FALSE OP-IS.GB-101 is V IGBHV-9531B-4 STC FALSE OP-IS.GS-101 052 V IGSHV-5050A STC 0 FALSE OP-IS.GS-101 056 V IGSHV-5052A STC 0 FALSE OP-IS.GS-101 070 V 1GSHV-5050B STC 0 FALSE OP-IS.GS-101 076 V IGSH-V-5052B STC 0 FALSE OP-IS.GS-101 086 V 1GSHV-5053B STC .0 FALSE OP-IS.GS-101 090 V 1GSHV-5054B STC 0 FALSE OP-IS.GS-101 112 V 1GSHV-5053A STC 0 FALSE OP-IS.GS-101 116 V IGSHV-5054A STC 0 FALSE OP-IS.GS-101 082 V IGSHV-4978 STC 0 FALSE OP-IS.GS-101 . 048 V 1GSHV-4952 STC" 0. FALSE OP-IS.GS-101 050 V 1GSHV-4951 STC 0 FALSE OP-IS.GS-101 001 V IGSHV-4955B STC 0 FALSE OP-IS.GS-101 005 V .IGSHV-4983B STC 0 FALSE OP-IS.GS-101 009 V 1GSIIV-5019B STC 0 FALSE OP-IS.GS-101 015 V IGSI{V-4984B STC .0 FALSE OP-IS.GS-101 040 V IGSHV-4955A STC 0 FALSE OP-IS.GS-101 044 V 1GSHV-4983A STC 0 FALSE OP-IS.GS-101 036 V 1GSHV-4984A STC 0 FALSE OP-IS.GS-101 120 V 1GSHV-4959A STC .0 FALSE OP-IS.GS-101 124 V 1GSHV-4965A STC 0 FALSE OP-IS.GS-101 136 V 1GSHV-4966A STC 0 FALSE OP-IS.GS-101 140 V 1GSHV-5022A STC 0 FALSE OP-IS.GS-101 060 V IGSHV-4974 STC 0 FALSE OP-IS.GS-101 092 V IGSHV-4963 STC . 0 FALSE OP-IS.GS-101 094

V 1GSHV-5029 STC 0 FALSE OP-IS.GS-101 128 V IGSHV-5031 STC 0 FALSE OP-IS.GS-101 104 V 1GSHV-4959B STC 0 FALSE OP-IS.GS-101 108 V 1GSHV-4965B STC 0 FALSE OP-IS.GS-101 062 V 1GSHV-4966B STC 0 FALSE OP-IS.GS-101 066 V 1GSHV-5022B STC 0 FALSE OP-IS.GS-101 078 V IGSHV-4956 STC 0 FALSE OP-IS.GS-101 080 V IGSH-V-4979 STC 0 FALSE OP-IS.GS-101 134 V IGSHV-4958 STC 0 FALSE OP-ISGS-101 038 V IGSHV-4950 STC 0 FALSE OP-IS.GS-101 098 V IGSHV-4962 STC 0 FALSE OP-IS.GS-101 100 V IGSHV-4964 STC 0 FALSE OP-IS.HB-101 1 V 1I-IBHV-F019 STC FALSE OP-IS.HiB-101 2 V 1HBHV-F020 STC FALSE OP-IS.HB-101 4 V IHBHV-F003 STC

FALSE OP-IS.H-IB- 101 .5 V 1HBHV-F004 STC FALSE CC-AA-309-1001, Rev 3

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO. 3 PAGE NO. 107 of 109 PROCEDURE STEP NO PV COMPONENT EXAM SEQUENCE D S OP-IS.KL-101 01 V 1KLHV-5147 STC 0 FALSE OP-ISKL-t01 05 V 1KLHV-5156A STC 0 FALSE OP-IS.KL-101 07 V 1KLHV-5126A STC .0 FALSE OP-IS.KL-101 13 V IKLH{V-5154 STC 0 FALSE OP-IS.KL-101 15 V 1KLHV-5155 STC 0 FALSE OP-IS.KL-101 19 V 1KLHV-5152A STC 0 FALSE OP-IS.KL-102 05 V IKLHV-5156B STC 0 FALSE OP-IS.KL-102 09 V 1KLHV-5126B STC 0 FALSE OP-IS.KL-102 13 V 1KLHV-5152B STC 0 FALSE OP-IS.KL-102 17 V 1KLHV-5124B STC 0 FALSE OP-IS.KL-103 I V 1KLHV-5148 STC _ FALSE OP-IS.KP-103 2 V IKPHV-6055B STC FALSE OP-IS.RC-101 1 V 1RCSV-643A STC FALSE OP-IS.RC-101 3 V 1RCSV-643B STC FALSE OP-IS.RC-101 5 V 1RCSV-707A STC FALSE OP-IS.RC-101 7 V 1RCSV-707B STC FALSE OP-IS.RC-101 9 V 1RCSV-728A STC FALSE OP-IS.RC-101 11 V IRCSV-728B STC FALSE OP-IS.RC-101 13 V 1RCSV-729A STC FALSE OP-IS.RC-101. 15 V IRCSV-729B STC FALSE OP-IS.RC-10.1 17 V 1RCSV-730A STC FALSE OP-IS.RC-101 19 V IRCSV-730B STC FALSE.

OP-IS.RC-101 21 V 1RCSV-731A STC FALSE OP-IS.RC-101 23 V IRCSV-731B STC FALSE OP-IS.SE-101 1 V 1SEHV-5161 STC FALSE OP-IS.SE-101 3 V 1SESV-J004-A1 STC FALSE OP-IS.SE-101 4 V 1SESV-J004-A2 STC FALSE OP-IS.SE-101 5 V lSESV-J004-A3 STC FALSE OP-IS.SE-101 6 V ISESV-J004-A4 STC FALSE OP-IS.SE-101 7 V ISESV-J004-A5 STC FALSE OP-IS.SK-101 1 V 1SKHV-5018 STC FALSE OP-IS.SK-101 3 V 1SKI-IV-4953 STC FALSE OP-IS.SK-101 5 . V 1SKHV-4957 STC FALSE OP-IS.SK-101 7 V ISKH-IV-4981 STC FALSE-OP-IS.BC-103 01 V IBCHV-FO04C STC 0 FALSE OP-IS.BC-103 03 V 1BCHV-F007C STC 0 FALSE OP-IS.BC-103 07 V IBCHV-F007A STC 0 FALSE OP-IS.EG-101 51 V IEGSV-2281-2 STC 0 FALSE OP-IS.EG- 101 53 V 1EGSV-2288-1 STC 0 FALSE OP-IS.EG-101 55 V 1EGSV-2288-2 STC 0 FALSE MD-GP.ZZ-100 2 V IKJSV-7534A STC FALSE MD-GP.ZZ-100 5 V IKJSV-7534B STC FALSE MD-GP.ZZ-100 8 V IKJSV-7534C STC FALSE MID-GP.ZZ-100 11 V 1KJSV-7534D STC FALSE

OP-IS.EA-102 1 V 1EAHV-2198D STC FALSE OP-IS.EA-102 5 V 1EAHV-2197B STC FALSE OP-IS.EA- 102 9 V IEAHV-2204 STC FALSE OP-IS.EA- 102 37 V 1EASV-2239 STC 0 FALSE OP-IS.EA-102 23 V IEAHV-2198B STC FALSE OP-IS.EA-102 27 V 1EAHV-2197D I STC _ FALSE CC-AA-309-1001, Rev 3

I CALCULATION NO. H-1-ZZ-MDC-1880 - REVISION NO. 3 j PAGE NO. 108 of 109 PROCEDURE STEP NO PV COMPONENT EXAM SEQUENCE D S OP-IS.EA-102 39 V 1BCSV-F074 STC 0 FALSE OP-IS.KJ-101 02 V 1KJSV-7534A STC 0 FALSE OP-IS.KJ-102 02 V 1KJSV-7534B STC 0 FALSE OP-IS.KJ-103 .02 V 1KJSV-7534C STC 0 FALSE OP-IS.KJ-104 02 V 1KJSV-7534D STC 0 FALSE OP-IS.RC-102 1 V 1RCSV-8903A STC FALSE OP-IS.RC-102 3 V IRCSV-8903B STC _ FALSE OP-IS.EA-102 31 V IECHV-4648 STC FALSE OP-IS.BD-101 15 V 1BDHV-F046 STC .0 FALSE OP-IS.BJ-101 33 V IFDHV-F026 STC 0 FALSE OP-IS.3J-101 41A V IBJHV-F004 STC 0 FALSE OP-IS.BG-102 01. V IBGHV-F001 STC 0 FALSE OP-IS.BG-102 03 V IBGHV-F004 STC 0 FALSE OP-IS.BG-102 05 V 1AEHV-F039 STC 0 FALSE OP-IS.AB-104 01 V 1ABHV-F022A STC FALSE OP-IS.AB-104 02 V 1ABHV-F022B STC FALSE OP-IS.AB-104 03 V IABHV-F022C STC FALSE OP-IS.AB-104 04 -7V IABHV-F022D STC FALSE CC-AA-309-1001, Rev 3

CALCULATION NO. H-1-ZZ-MDC-1880 REVISION NO. 3 PAGE NO. 109 of 109 Attachment 14.3 1 CD With Various Electronic Files I CC-AA-309-1001, Rev 3 1

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10.0 REFERENCES

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