RS-05-114, Rev. 0 to Calculation DRE05-0048, Dresden Units 2 & 3 Post-LOCA Eab, LPZ, and CR Dose - AST Analysis.

Rev. 0 to Calculation DRE05-0048, Dresden Units 2 & 3 Post-LOCA Eab, LPZ, and CR Dose - AST Analysis.

ML052430393
Person / Time
Site:	Dresden
Issue date:	08/14/2005
From:	Gita Patel Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
FOIA/PA-2010-0209, RS-05-114 DRE05-0048, Rev 0
Download: ML052430393 (65)

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ATTACHMENT 8 Calculation DRE05-0048, "Dresden Units 2 & 3 Post-LOCA EAB, LPZ, and CR Dose - AST Analysis," Revision 0

CC-AA-309-1001 Revision 2 ATTACH1IMENT 1 Desihn Analysis Cover Sheet Design Analysis (Major Revision) Last Page No. I 64 Analysis No.: DREO5-0048 Revision: ' 0

Title:

' Dresden Units 2 & 3 Post-LOCA EAB, LPZ, and CR Dose - AST Analysis ECIECR No.:' 366383 Revision: 0 Station(s): ' Dresden Component(s):

Unit No.:' 00 NIA Discipline: ' MECH Descrip. Code/Keyword:" AST Safety/QA Class: " SR System Code: 12 00 Structure: " W/A CONTROLLED DOCUMENT REFERENCES _

Document No.: From/To Document No.: rom/To TODI No. ER2002-9994, Rev I From For remaining References see From GE-NE-A22-00103-64-01, Rev 0 From Section 9.0 GE-NE-A22-00103-08-01, Rev 1 From DREO4-0030, Rev 1 From DRE02-0033, Rev 0 From DRE01-0040, Rev O From __

Isthis Design Analysis Safeguards Information?" Yes ] No 0 If yes, see SY.AA-101-106 Does this Design Analysis contain Unverified Assumptions? Yes 0 No ED f Yes, AT/AR#:

This Design Analysis SUPERCEDES: DRE01-0040, Rev. 0 In Its entirety.

Description of Revision (list affected pages for partials): " New Design lysis Preparer. Gophl Paul (NUCORE) . _ _ _ _._ - 08114/05 Priniem _Sign_

Method of Review.: Detailed Review 03 Alternate CalculaI ffached) Testing O Reviewer Mark Drucker (NUCORE) 3Ak - .. 0811410S sPf~tdM ~n Moe Mm".

Review Notes: 2 Independent review JR Peor review 0 4SAA:T &e-o~~'4 I ClCl- .rAZt 0 - dv:

ijw tt:~~~~,

4A 14 ,, .TZ External Approver 3 Gopal Patel ,* 124l1 Exelon Reviewer:" Thomas Mscisa ",,

Is a Supplemental Review Required? Yes NobA if a ;oN teeAttachment 3 Exeion Approver: v Elliott Flick . -. (L.-'-o Py "

This Page Is Intentionally Left Blank CC-AA-309-1001 Rev. 2

ATTACIIENT 2 Owners Acceptance Review Cieeldist for External Design Analysis Page 1 of I DESIGN ANALYSIS NO. DRE05-0048 REV: 0 Yes No N/A

1. Do assumptions have sufficient rationale? 0 0 Are assumptions compatible with the way the plant is operated and with the licensing El

2. basis? 0 0

3. Do the design inputs have sufficient rationale? 0 03 0

4. Are design inputs correct and reasonable? 0 0o 01 5 ' Are design inputs compatible with the way the plant is operated and with the 11 licensing basis? 0 0

6. Are Engineering Judgments clearly documented and justified? 03 01 11 7 Are Engineering Judgments compatible with the way the plant is operated and with the licensing basis? ED 0l 11

8. Do the results and conclusions satisfy the purpose and objective of the Design Analysis? 0E 0 0

9. Are the results and conclusions compatible with the way the plant is operated and with the licensing basis? 0 0 0

10. Does the Design Analysis include the applicable design basis documentation? 0 0 .0 Have any limitations on the use of the results been identified and transmitted io the IL appropriate organizations? 0 0

12. Are there any unverified assumptions? 0 0 .0

13. Do all unverified assumptions have a tracking and closure mechanism in place? 0 0 0 Have all affected design analyses been documented on the Affected Documents List

14. (ADL) for the associated Configuration Change? 0 0 0 Do the sources of inputs and analysis methodology used meet current technical requirements and regulatory commitments? (If the input sources or analysis

• 5 methodology are based on an out-of-date methodology or code, additional 0 0 0 reconciliation may be required if the site has since committed to a more recent code)

16. Have vendor supporting technical documents and references (including GE DRFs) been reviewed when necessary? 0 0 0 EXELON REVIEWER: Thomas Mscisz Print / Sign DATE: 4 6 CC-AA-309-1001 Rev. 2

ICALCULATION NO. DRE0S-0048 IREVISION NO. 0 7 PAGE NO. 4 of 64 l REVISION HISTORY Revision I Issue Date Revision Description l 0 Original Issue I CC-AA-309-001.,Rev22

1 I PAGE NO. 5 of 64 CALCULAT1ON NO. DREO5-0048 - l REVISION NO.0 I PAGE REVISION INDEX PAGE REV PAGE REV I 0 36 0 2 0 37 0 3 0 38 0 4 0 39 0 5 0 40 0 6 0 41 0 7 0 42 0 8 0 43 0 9 0 44 0 10 0 45 0 11 0 46 0 12 0 47 0 13 0 48 0 14 0 49 0 15 0 so 0 16 0 51 0 17 0 52 0 18 0 53 0 19 0 54 0 20 0 55 0 21 0 56 0 22 0 57 0 23 0 58 0 24 0 59 .0 25 0 60 0 26 0 61 0 27 0 62 0 28 0 . . 63 0 29 0 64 0 30 0 31 0 32 0 33 0 34 0 35 0 I CC-AA-309-1001, Rev2

CALCULATION NO. DRE05-0048 _RENFISION-NO.0 -PAGE NO. 6 of 64 TABLE OF CONTENTS Section Sheet No.

Design Analysis Cover Sheet 1 Revision History 4 Page Revision Index 5 Table of Contents 6 1.0 Purpose 7 2.0 Methodology 7 3.0 Acceptance Criteria 14 4.0 Assumptions 15 5.0 Design Inputs 19 6.0 Computer Codes & Compliance With Regulatory Requirements 25 7.0 Calculations 26 8.0 Results Summary & Conclusions 38 9.0 References 39 10.0 Tables 41 11.0 Figures 59 12.0 Affected Documents 64 13.0 Attachments 64 CC-AA-309-1001, Rev 2

_ CALCULATION NO. DRE0-0048 REVISION NO. 0 -J PAGE NO. Tof -64 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 the Dresden Nuclear Power Station (DNPS) using as-built design inputs and assumptions, the Alternative Source Term (AST), the guidance in Regulatory Guide (RG) 1.183, and Total Effective Dose Equivalent (TEDE) dose criteria.

This calculation is performed in a reasonably conservative manner in which 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.

2.0 METHODOLOGY AND ACCEPTANCE CRITERIA The design basis loss of coolant accident is analyzed using a conservative set of assumptions and as-built design inputs parameters compatible for the AST and TEDE dose criteria. The numeric values of the critical design inputs are conservatively selected to assure an appropriate 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 Leakaze 2.1.1 Source Term The post-LOCA containment leakage model is shown in Figure 1. The BWR core inventory fractions

' listed in Regulatory Guide 1.183 Table I are released into the containment at the release timing shown in RG 1.183 Table 4 (Ref. 9.1, Sections 3.2 & 3.3). Since the post-LOCA minimum suppression chamber water pH is greater than 7.0 (Ref. 9.12), 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 (Ref. 9.1, Section A.2). With the exception of elemental and organic iodine and noble gases, the remaining fission products are assumed to be in particulate form (Ref. 9.1, Section 3.5). The isotopic core inventory (Ci/MWt) of fission products in the reactor core is obtained from Reference 9.6, Appendix D and listed in Design Input (DI) 5.3.1.3. The RADTRAD Nuclide Inventory File (NIF)

DPSdef.TXT is developed based on the plant-specific core inventory and used for the containment, ESF, and MSIV leakage paths. The source term design inputs are shown in Sections 5.3.1.1 through 5.3.1.7.

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 as it is released 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. 9.1, Table 4). The design inputs for the transport in the primary containment are shown in Sections 5.3.2.1 through 5.3.2.11. The reduction in containment leakage activity by dilution in the RB and removal by the SBGTS filtration are credited.

CC-AA-309-1001, Rev 2

l CALCULATION NO. DRE05-0048 -REVISIONNO. 0 lPAGE NO. 8 of 64 The analysis dilutes the radioactivity released from the core into the drywell volume during the first 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of the LOCA, and then into the combined drywell plus suppression chamber air volume after 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, 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 of steaming and condensing phenomenon (Ref. 9.5, Table 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 txhe aerosols from the containment. Iodine removal by suppression' pool scrubbing is not credited because the bulk core activity is release to containment well after the initial mass and energy release. Although containment sprays are not credited, the removal of the elemental iodine by deposition on surfaces inside containment is modeled in the same way as containment spray. The DF of elemental iodine is based on the SRP 6.5.2 guidance and is limited to a DF of 200 (see Section 7.7) (Ref. 9.9, page 6.5.2-10). The containment leakage of 0.03 volume fractions per day (i.e., 3 volo/day) is assumed, which is the sum of the primary-to-secondary leakage and leakage through the MSIVs. Reduction in the containment leakage afte 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is not credited in the analysis.

2.1.4 Dual Containment Leakage from the primary containment is assumed to mix in 50% of the reactor building (RB) free air volume. The 50% mixing effectively reduces the RB net free volume by 50% when modeled for the containment & ESF leakage releases.

2.1.5 Containment Purging The containment purging during a LOCA is not a credible event for the DNPS (Ref. 9.4, Item 3).

Therefore, the release from containment purging is not analyzed per RG 1.183, Section A.7.

2.2 Post-LOCA ESF Leakaue The post-LOCA ESF leakage release model is shown in Figure 1. 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 are analyzed and combined with the radiological consequences from other fission product release paths to determine the total calculated radiological consequences from the LOCA (see Section 8.1 of this calc). The ESF components are located in the RB.

2.2.1 Source Tenn 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 systems is assumed to be 1 gpm. This ESF leakage is doubled (Ref. 9. 1, Section A.5.2) and assumed to start at time t = 0.0 minutes after the 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.

CC-AA-309-1001, Rev 2

l CALCULATION NO. DRE05-0048 _l REVJSION NO.0 - PAGE NO. 9 of 64 Since the temperature of the recirculating liquid is less than 212'F, 10% iodine activity in the ESF is assumed to become airborne (Ref. 9.4, Item 29). The design inputs for the ESF leakage are shown in Section 5.4. The reduction in ESF leakage activity by dilution in 50% of the RB volume and removal by the SBGTS filtration are credited.

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 9.1, Section A.5.6).

2.3 Post-LOCA MSIV Leakage The post-LOCA MSIV Leakage model is shown in Figure 2. The four main steam lines, which penetrate the primary containment, are automatically isolated by the MSIVs in the event of a LOCA (Ref. 9.15).

There arc two MSIVs on each steam line, one inside containment and one outside containment. The MSIVs are finctionally part of the primary containment boundary and design leakage through these valves provides a leakage path for fission products to bypass the'secondary containment and enter the environment as a ground-level release. Following the initial blowdown of the reactor pressure vessel (RPV), the steaming in the RPV carries fission products to the containment. When core cooling is restored, the steam and the ESF flow carry fission products from the core to the primary containment via the severed recirculation line, resulting in well-mixed RPV dome and containment fission product concentrations. The main steam isolation valves (MSIVs) are postulated to leak at a total design leak rate of 150 scfl. 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 acceptable 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 activity released into the drywell for evaluating containment leakage.

A total of 150 scfh MSIV leakage is assumed to occur as follows (see Section 2.3.2 for additional information regarding steam line selection):

(1) 60 scfh through the steam line with the failed inboard MSIV. Conservatively, the deposition of aerosol and elemental iodine activities are not credited in the steam line between the RPV nozzle and the failed inboard MSIV. The deposition removal of aerosols and elemental iodine is credited in the horizontal pipe between the inboard and outboard MSIVs.

(2) 60 scfh through first intact steam line. The deposition removal of aerosol in the horizontal pipe, and the deposition removal of elemental iodine in both the horizontal and vertical pipes, are credited in the steam line between the RPV nozzle and outboard MSIV.

(3) 30 scfh through second intact steam line. The deposition removal of aerosol in the horizontal pipe, and the deposition removal of elemental iodine in both the horizontal and vertical pipes, are credited in the steam line between the RPV nozzle and outboard MSIV.

(4) 0 scfh through the fourth steam line.

CC-AA-309-1001,Rev2 I

CALCULATION NO. DRE05-0048 _l REVISION NO.0 _lPAGE NO.I0 of64 The aerosol deposition removal efficiencies for the main steam lines are determined based on the methodology in Appendix A of AEB-98-03 (Ref. 9.5) using only the horizontal pipe projected area (Diameter x Length) as shown in Table 2. 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.

This treatment of elemental iodine includes the resuspension and fixation of elemental iodine from the pipe surface.

2.3.2 Determination of MSIV Leak Rates In Various Steam Line Volumes The horizontal lengths of the three shortest steam headers in the Dresden plant (Rcfs. 9.15 & 9.16) are compared with those in the Quad Cities Nuclcar Power Station (QCNPS) (Ref. 9.20, Section 7.3) to determine if the aerosol deposition filter efficiencies established for the QCNPS MSIV leakage release path analysis can be used for the Dresden MSIV leakage release path analysis. A comparison of the total horizontal lengths between the reactor pressure vessel and the outboard MSIV of the three shortest main steam piping runs of QCNPS, DNPS Unit 2 and DNPS Unit 3 is shown in Table 1. It can be seen from the comparison that the QCNPS main steam piping runs are shorter, and would therefore result in less (i.e., conservative) aerosol deposition. Therefore, the settling velocities in the different steam lines (Table 2), the time dependent MSIV leak rates (Table 3), and the aerosol removal efficiencies (Table 4) are obtained from Reference 9.20 and conservatively used for the MSIV leakage analysis for the DNPS in the following section with the appropriate DNPS vs. QCNPS plant-specific changes.

The total MSIV leakage from all main steam lines is proposed to increased from 79.6 scfh to 150 scfh measured at 48 psig, allowing a maximum of 60 scfh from any one of the 4 main steam lines. The total MSIV leak of 150 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 outboard MSIVs are conservatively calculated in Section 7.2 using the Ideal Gas Law and drywell post-LOCA peak pressure and temperature and listed in Table 3. The MSIV leak rates in Table 3 are used in the analysis with aerosol removal efficiencies calculated in Table 4 based on the horizontal pipe surface areas calculated in Section 7.3. The reduction in the containment leakage and MSIV leakage after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of onset of a LOCA is not credited in the analysis. To account for the assumed mixing between the wetwell and drywell at 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 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.

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 response. Therefore, a recirculation line rupture is considered as the initiating event rather than a main steam line rupture.

Per UFSAR Section 6.2.1.3.5.1, the DBA for the Mark I containment design is the instantaneous guillotine rupture of the largest pipe in the primary system (the recirculation suction line). This LOCA leads to a specific combination of dynamic, quasi-static, and static loads in time. The thermal transient CC-AA-309-1001, Rcv2

l CALCULATION NO. DREOS-0048 4

..REuS]ON NO. 0 -I-PAGE-NO. 11 of-64 I due to other postulated events including the steam line break inside the drywell does not impose maximum challenge to drywell pressure boundary and fuel integrity. The LOCA results in the maximum core damage and fission product release as shown in the RG 1.183 (Ref 9.1, Table 1). Therefore, a recirculation line rupture is considered to be the limiting event with respect to radiological consequences.

RG 1.183 (Ref. 9.1, Appendix A, Section 6.5) allows reduction in MSIV releases that is due to holdup

---

and-deposition in main-steamnpiping-downstream of the MSIVs and in the main condenser, including the treatment of air ejector effluent by offgas systems, if the components and piping systems used in the release path are capable of performing their safety function during and following a safe shutdown earthquake (SSE). Although postulating a main steam line break in one steam line inside the drywell would maximize the dose contribution from the MSIV leakage, the steam line break is not a credible event during a LOCA, since the ASME Category I main steam piping is designed to withstand the SSE.

However, this analysis assumes that a steam line with MSIV failure does not credit removal of elemental or aerosol iodine in the piping between the reactor pressure vessel nozzle and inboard MSIV for conservatism.

2.4 Control Room Model The shielding analysis for CR operator exposure from various sources is performed in the following sections using the best available information from Exelon Engineering, and drawings provided by Exelon at the time of analysis. The shielding information is used in a reasonably conservative manner.

The post-LOCA control room RADTRAD nodalization is shown in Figure 3 with the design input parameters. The post-LOCA radioactive releases that contribute to the CR TEDE dose are as follows:

• Post-LOCA Containment Leakage

• Post-LOCA ESF Leakage Post-LOCA MSIV Leakage The radioactivity from the above sources is 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 filtration system prior to being distributed in the CR envelope. The 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 contairanent shine to CR

• Post-LOCA Control Room Emergency Ventilation (CREV) filter shine 2.4.1 Post-LOCA Airborne Activity Inside CR The post-LOCA radioactive releases from various sources are shown in Figures I and 2. The activities released from the various sources are diluted by atmospheric dispersion and carried to the CR air intake.

The atmospheric dispersion factors are shown in Sections 5.6.8 & 5.6.9 for the containment/ESF and MSIV leakages. The containment and ESF leakages have the same release point (Station Chimney) and X/Qs. The RADTRAD release models are developed for each release path using appropriate design inputs from Sections 5.3 through 5.5. The CR dose model is developed using the design input CC-AA-309-1001. Rev 2

4 l[CALCULATION-NO-DRE05-0048 . ___ lREVISIONNO.0 --- l-PAGE NO.12 of 64 .

parameters in Section 5.6. The CR airborne TEDE dose contributions from the above post-LOCA sources are calculated and tabulated in Section 8.1.

2.4.2 Post-LOCA Airborne Cloud External to CR The post-LOCA radioactive plume contains the radioactive sources from the containment, ESF, and MSIV leakages. The gamma radiation external radioactive plume shine to the CR personnel is attenuated by the 1'-6" minimum concrete wall shielding (Ref. 9.22.c). The RADTRAD3.03 code calculates the whole body gamma dose based on the semi-infinite cloud immersion at site boundary location (Ref. 9.2, Section 2.3.1 and Ref. 9.1, Section 4.1.4). Therefore, the X/Qs for the LPZ receptor modeled in RADTRAD file DRE400MS31.psf are modified by replacing them with the X/Qs for the CR air intake locatiop. Since the containment and ESF leakages contribute insignificant CR doses (Section 8.0), they are not considered important sources for the external cloud dose. The resulting LPZ whole body dose is the semi-infinite gamma dose at the CR air intake. The total whole body gamma dose is 15.89 rem, which is obtained from RADTRAD run DRE400MS32.oO. Since this is a semi-infinite dose at the CR air intake, it is appropriate to assign this dose to the CR roof. The gamma attenuation factor is calculated in Section 7.6 to be 0.0172 for a I Mev gamma emission. This attenuation factor includes the buildup due to multiple scattering. The resulting gamma dose from the external cloud shine would be 0.273 rem (15.89 rem x 0.0172 = 0.273 rem), which is added with the dose contribution from other post-LOCA sources in Section 8.1.

2.4.3 Post-LOCA Containment Shine to CR The CR location with respect to the reactor building is shown in Reference 9.21. The post-LOCA airborne activity in the containment (drywell) 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 space above the operating floor of the RB (Ref. 9.2 l.d) contributes direct shine dose to the CR operator. The 0-24 hrs post-LOCA RB airborne activity from the containment leakage and ESF leakage are listed in Tables 5 & 6 and combined in Table 7. The review of the containment building concrete structure drawing (Ref. 9.22) indicates the concrete walls near Column 33 total 2'-4" thick (P'-6" + 10" = 2'-4') (Ref. 9.22.b). The line of sight from the CR operator location to the RB source involves multiple shadow shields consisting of the concrete roof & walls and equipment. Therefore, the concrete shielding of 2'-6" is credited in the shielding model. Actually, a large shadow concrete shielding will interact with gamma dose direct line of sight from the CR operator to the RB operating floor as shown Figure 4 (Refs. 9.21 & 9.22). The CR operator direct shine dose is dependent on the shielding geometry between the CR and RB and post-LOCA source term. The shielding configuration (line of sight distance and intercepting concrete shielding) for the Dresden CR operator dose are virtually identical to Quad Cities. For the given shielding geometry, the CR dose is dependent on the RB source term. The comparison of the total isotopic activities in Table 7 with the corresponding isotopic activities in the Quad Cities RB in Reference 9.20, Table 6 indicates that the isotopic activities in the Quad Cities RB are slightly higher than those in the Dresden RB. Therefore, the post-LOCA containment shine dose of 0.22 rem (Ref. 9.20, Sections 7.8 & 7.9) calculated for the Quad Cities CR operator can be conservatively applied to the Dresden CR operator. The resulting containment shine dose is listed in Section 8.10.

2.4.4 Post-LOCA CREV Filter Shine The Dresden combined CR is located at the east end of the plant in the service building between Rows D

& H, adjacent to Column 33 (Ref. 9.22). The CREV charcoal filter Tag number 9400-101 was obtained from a general arrangement drawing (Ref. 9.23.a) to locate the filter on the HVAC drawings (Ref. 9.23).

CC-AA-309-1001,Rev2

j CALCULATION NO. DRE05::0048 __l.REYISION-NO.0 . _TPAGENO.-I3or64 The CREV charcoal filter is located in the south-west comer of the service building at EL 534'-O" (Ref.

9.23) near the intersection of Row H and Column 31. The CREV charcoal filter is located south of the CR. The dimensions of charcoal filter housing are approximately 3'-3" x 5'3" x 16'-6" (Ref. 9.24). The post-LOCA CR doses listed in Section 7.0 indicate that the containment and ESF leakage contribute insignificant dose to the CR operator due to the' large atmospheric dilution provided by their elevated releases from the SBGT station chimney. Therefore, only the MSIV leakage path is used to assess iodine and aerosol activity on the CR charcoal filter in the following section.

The RADTRAD3.03 code calculates the cumulative elemental and organic iodine atoms and the aerosol mass released to the'environment from the main steam lines due to MSIV leakage at various time steps.

The activity released to the environment is atmospherically dispersed to the control room HVAC intake louvers, where it is drawn into the CREV System. Section 7.11 and Tables 9 through 14 calculate the total elemental and organic iodine atoms and aerosol mass drawn into, and retained on, the CREV charcoal and HEPA filters. Section 7.11 conservatively neglects decay of the isotopes deposited on the CREV filters.

2.4.4.1 Post-LOCA Iodine Activity On CR Charcoal Filter-MSIV Leakage The iodine atom/curie relationship is established using the containment leakage run DRE400CL3 I.oO file as shown in Table 15, which is a typical relationship for all release paths. The total number of atoms accumulated on the charcoal filter is established in Section 7.11 based on the charcoal filter efficiency and CREV intake flow rate. Knowing the iodine atom/curie relationship (Table 15), and the total number of elemental and organic iodine atoms on the charcoal filter (Tables 10 and 12), the total (elemental + organic) iodine activity deposited on the CREV charcoal filter due to the MSIV leakage is calculated in Section 7.11 (Table 16). The review of Table 16 indicates the accumulation of iodine is insignificant. This is as expected, because most of the elemental iodine is removed by elemental deposition in the main steam piping before it is released to the environment and it is further reduced by' air dilution before it migrates to the CR air intake.

2.4.4.2 Post-LOCA Aerosol Activity On CR HEPA Filter - MSIV Leakage The aerosol mass/curie relationship is established using the containment leakage run DRE400MS3 l.oO file as shown in Table 17, which is a typical relationship for all release paths. The total aerosol mass deposited on the CREV HEPA filter due to the MSIV leakage is calculated in Section 7.11 based on the HEPA filter efficiency and CREV intake flow rate. Knowing the aerosol mass/curie relationship (Table 17), and the total mass of aerosols on the HEPA filter (Table 14), the total aerosol activity deposited on the CREV charcoal filter due to the MSIV leakage is calculated in Section 7.11 (Table 18). The isotopic aerosol activity in Table 18 is insignificant. This is as expected, because most of the aerosols deposit out in the main steam piping horizontal surface before being released to the environment (see Table 4 for the aerosol removal efficiencies due to gravitational deposition).

2.4.4.3 Concrete Shielding With CREV Charcoal Filter The CREV charcoal filter is located in the south-west corner of the service building at EL 534'-0" (Ref. 9.23) near the intersection of Row H and Column 31. The CREV charcoal filter housing is at least 4.75' from the control room:

[(Distance between Rows Hl and H + Distance between Row H & south wall of CR + Thickness of CR south wall] - [(Distance between the Column HI & south edge of filter housing + length of filter housing)]

CC-AA.309-1001, Rev 2

CALCULATION NO. DREOS-0048 REVISION NO. 0 l PAGE NO. 14 of 64

= [(9'-6") (Ref. 9.22.b) + (14'-3') (Ref. 9.22.b) + 1'-9"] - [4'-3" (Ref. 9.23.a) +16'-6" (Ref. 9.24)]

=25'6"- 20'-9" = 4.75' The line of sight between the CR operator location and the CREV filter is mainly intercepted by the 1'-9" concrete wall located at south of CR. (Refs. 9.23.b & 9.23.c). In addition, shadow shielding is afforded by the equipment and duct work on the floor. The post-LOCA iodine and aerosol sources are small (Tables 16 & 18) as discussed in Sections 2.4.4.1 & 2.4.4.2 above, which coupled with the large amount of concrete shielding that exists between the CR operator and the CREV charcoal filter, makes the CREV charcoal filter shine dose insignificant to the CR operator.

3.0 ACCEPTANCE CRITERIA The following NRC regulatory requirement and guidance documents are applicable to this DNPS Alternative Source Term LOCA Calculation:

0 Regulatory Guide 1.183 (Ref. 9.1) 0 IOCFR50.67 (Rcf. 9.3) 0 Standard Review Plan section 15.0.1 (Ref. 9.25)

Dose Acceptance Criteria are:

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

TEDE Dose I 5 - 25 CC-AA-309-1 001, Rev22

1- CALCULATIONNO. DRE05-0048 lVEMISIONNO. 0 [PAGENO.15 of 64 4.0 ASSUMPTIONS The following assumptions used in evaluating the offsite and control room doses resulting from a Loss of Coolant Accident (LOCA) are based on the requirements in the Regulatory Guide 1.183 (Ref. 9.1).

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 Guide Positions (RGPs) 3.1 through 3.4 of Reference 9.1 as follows:

4.2 Equilibrium 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 3,016 MWt, which represents the maximum full power operation of the core at a power level equal to the Extended Power Uprate (EPU) thermal power level of 2,957 MWt plus a 2% margin for instrument uncertainty (Ref. 9.4, Item 1). The equilibrium core inventory is described in Design Input 5.3.1.3.

4.3 Release Fractions and TimLnj The core inventory release fractions, by radionuclide groups, for the gap release and early in-vessel damage for a Design Basis Accident (DBA) LOCA' arc listed in Design Input 5.3.1.5. These fractions are applied to the equilibrium core inventory (Ref. 9.1, Tables 1 & 4). The release fractions are acceptable for use given that the peak fuel burnup meets the 62,000 MWD/MTU requirement specified in Regulatory Guide 1.183 (Ref. 9.1,Note 10).

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. 9.1, RGP 3.4).

4.5 Chemical Form

'The long-term suppression pool water pH is greater than 7 during a LOCA (9.12, page 11) with credit taken for sodium pentaborate in the Standby Liquid Control System. 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. 9.1, RGPs 3.5 and A.2). These are shown in Design Input 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. 9.1, RGPs 3.5 and A.2).

4.6 Assumptions on Activity Transport in PrimarV 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. The radioactivity released from the fuel doesn't mix with the suppression pool air space until after two hours, as previously discussed in Section 2.3.2.

4.6.2 Reduction in airborne radioactivity in the containment by natural deposition within the containment is credited using the'RADTRAD3.03 Powers model for aerosol removal coefficient with a 10-percentile probability (Ref. 9.1, RGP A.3.2; & Ref. 9.2, 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 event duration (Ref. 9.1, RGP A.3.7).

CC-AA-309-1001. Rev2

CALCULATION NO. DRE0S-0048 REISION NO. 0 l PAGE NO. 16 of i 4.6.4 The Dresden Station does not purge containment to relieve containment pressure or to reduce containment hydrogen concentration (Ref. 9.4, Item 3). Therefore, the release from containment purging is not analyzed.

4.7 OfMsite 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 as a TEDE, which is the sum of the committed effective dose equivalent (CEDE) from inhalation and the deep dose equivalent (DDE) from external exposure from all radionuclides that are significant with regard to dose consequences and the released radioactivity (Ref. 9.1, RGP 4.1.1; and Refs. 9.7 & 9.8). The RADTRAD3.03 computer code (Ref. 9.2) performs this summation to calculate the TEDE.

4.7.2 The offsite dose analysis uses the Committed Effective Dose Equivalent (CEDE) Dose Conversion Factors (DCFs) for inhalation exposure. (Ref. 9.1, RGP 4.1.2; and Refs. 9.7 & 9.8).

4.7.3 Since RADTRAD3.03 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 EDE in lieu of DDE Dose Conversion Factors in determining external exposure (Ref. 9.1, RGP 4.1.4; and Ref. 9.8).

4.7.4 The maximum EAR TEDE for any tvo-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. 9.1, RGP 4.1.5 & RGP 4.4; and Ref. 9.3).

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 (Ref. 9.1, RGPs 4.1.6 and 4.4; and Ref. 9.3).

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. 9.1, RGP 4.1.7).

4.7.7 The breathing rates used for persons at offsite locations is given in Reference 9.1, RGPs 4.1.3 &

4.4. These rates are incorporated in Design Inputs 5.7.3 & 5.7.6.

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. 9.1, RGP 4.2.1). See applicable Design Inputs 5.6.1 through 5.6.11.

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-,-I_ CALCULATIONNO-DREOS-0048- _ 4

.REVISION NO.0 - -PAGE-NO.-17 of-64 I

• 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 containment shine from radioactive material in the reactor containment, and

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

4.8.2 The radioactivity releases and radiation levels used for the control room dose are determined using the same source term, transport, and release assumptions used for determining the exclusion area boundary (EAB) and the low population zone (LPZ) TEDE values (Ref. 9.1, 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 5.6.10 & 5.6.11 (Ref. 9.1, RGP 4.2.6).

4.8.4 10 CFR 50.67 (Ref. 9.3) 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. 9.1, 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 filtration (Ref. 9.1, 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 40 minutes (Design Input 5.6.2) after a LOCA (refer to Figure 3).

4.8.6 The CR unfiltered inleakage is conservatively assumed to be 2000 cfm during normal mode of CR HVAC operation (Design Input 5.6.5), and 400 cfm during emergency mode of CR HVAC operation (Design Input 5.6.6). The CR unfiltered inleakage of 2,000 cfm with normal flow intake flow.rate of 2,200 cfm is considered in the analysis during 040 minutes to justify the potential unfiltered leakage during the normal mode of operation. The unfiltered inleakage of 2,000 cfln represents at least 7 times the maximum measured unfiltered inleakage of 253 cfm

• (162+/- 91 cfm) by Tracer Gas Testing during emergency mode. The modeled unfiltered CC-AA-309-1001, Rev 2 1

l CALCULATION NO. DRE050048 l REVISION NO. 0 PAGE NO. 18 of 64 inleakage rates include ingress/egress inleakage of 10 cfm. The atmospheric dispersion factors generated for the CR intake are representative for control room inleakage.

4.8.7 No credits for KI pills or respirators are taken (Ref 9. 1,RGP 4.2.5).

CC-AA-309-1001, Rev 2 l

CALCULATION 0 NO. DREOS-0048 J REVISION NO. 0 ] PAGE NO. 19 of 64 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 Dresden Station 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 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, arc 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 failures modeled in this calculation are the MSIV in one main steam line failing to close and the operation of the CREV system failing to start by Safety Injection signal.

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 a 10% higher flow rate for the CR Normal Operation air intake, use of a 10% lower flow rate for the CR Emergency Ventilation Mode air intake, 40 minutes delay in the CR Emergency Ventilation Mode initiation time, and use of ground release X/Qs demonstrate the inherent conservatisms in the plant design and post-accident response.

5.1.4 Meteorologv Considerations Atmospheric dispersion factors (X/Qs) for the onsite release points such as the Standby Gas Treatment System (SBGTS) stack for containment and ESF leakage release path and the edge of the MSIV room for the MSIV leakage release path are developed (Ref. 9.11) using the NRC sponsored computer code ARCON96. The EAB and LPZ X/Qs are developed using the Dresden Station plant specific meteorology and appropriate regulatory guidance (Ref. 9.11).

5.2 Accident-Specific Desian Inputs/Assumptions The design inputs/assumptions utilized in the EAB, LPZ, and CR habitability analyses are listed in the following sections. The design inputs are compatible with the requirements of the AST and TEDE dose criteria and the assumptions are consistent with those identified in Regulatory Position 3 and Appendix A of RG 1.183 (Ref. 9. 1). The design inputs and assumptions in the following sections represent the as-built design of the plant.

CC-AA-309-1001, Rev 2

.___ CALCULATION NO. DRE05-0048- _- _

-j -VONO. 0.

REVISION NO.0

... ..._.- __ PAGE NO. 20 of 64 PAGENO.2Oof 64 -l-Design Input Parameter Value Assigned Reference 5.3 Containment Leakage Model Parameters 5.3.1 Source Term 5.3.1.1 Thermal Power Level 3,016 MWt (includes 2% margin) 9.4, Item I 5.3.1.2 Peak Fuel Burnup 62,000 MWD/MTU 9.4, Items 5 and 6

____ __9_ _ _ RGP 3.2,-note 10 5.3.1.3 Isotopic Core Inventory (C MWt) (Ref. 9.4, Item 2; and Ref. 9.6, Appendix D)

Isotope Cl/MW, Isotope CI/NIW, Isotope ClUMW, CO-58' 1.529E+02 RU-103 431 IE+04 CS-136 2.379E+03 CO-60* 1.830E+02 RU-105 3.034E+04 CS-137 4.928E+03 KR-85 4.364E+02 RU-106 1.837E+04 BA-139 _ 4.888E+04 KR-85M 6.772E+03 RH-105 2.882E+04 BA-140 4.714E+04 KR-87 1.291E+04 SB-127 2.999E+03 LA-140 5.055E+04 KR-88 1.815E+04 SB-129 8.877E+03 LA-141 4.447E+04 RB-86 7.096E+01 TE-127 2.986E+03 LA-142 4.286E+04 SR-89 2.428E+04 TE-127M 4.060E+02 CE-141 4.465E+04 SR-90 3.528E+03 TE-129 8.735E+03 CE-143 4.101 E+04 SR-91 3.081E+04 TE-129M I.300E+03 CE-144 3.682E+04 SR-92 3.362E+04 TE-131M 3.9S5E+03 PR-143 3.963E+04 Y-90 3.625E+03 TE-132 3.850E+04 ND-147 1.800E+04 Y-91 3.155E+04 1-131 2.710E+04 NP-239 5.587E+05 Y-92 3.377E+04 1-132 3.914E+04 PU-238 1.768E+02 Y-93 3.942E+04 1-133 5.501 E+04 PU-239 1.474E+O1 ZR-95 4.443E+04 1-134 6.035E+04 PU-240 2.001 E+O I ZR-97 4.497E+04 1-135 S.157E+04 PU-241 6.700E+03 NB-95 4.464E+04 XE-133 5.282E+04 AM-241 9.857E+O0 MO-99 5.121E+04 XE-135 2.144E+04 CM-242 2.285E+03 TC-99M 4A84E+04 CS-134 8.009E+03 CM-244 1.621 E+02 C0-58 & CO-60 activities are obtained from RADTRAD User's Manual, Table 1.4.3.2-3 (Rcf. 9.2) 5.3.1.4 Radionuclide Composition

_Group Elements-Noble Gases Xe, Kr 9. 1, RGP 3.4, Table 7 Halogens I, Br_-

Alkali Metals Cs, Rb Tellurium Group Te, Sb, Se Barium, Strontium 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 CC-AA-309-1001, Rev 2

CALCULATION NO. DRE05-0048 _l.REVISION NO. 0 'PAGENO.21 of. 64 Design Input Parameter Value Assigned Reference 5.3.1.5 Release Fraction (Ref 9.1, Table 1)

BU'R 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 5.3.1.6 Timing of Release Phase LRef. 9.1, Table 4)

Phase Onset Duration Gap Release 2 min 0.5 hr Early In-Vessel Release 0.5 hr 1.5 hr 5.3.1.7 Iodine Chemical Form Aerosol (CsI) 95% 9.1, RGP 3.5 Elemental 4.85%

Organic 0.15%

5.3.2 Activity Transport in Prim ry Containment 5.3.2.1 Drywell Air Volume 158,000 fet 9.4, Items 13 and 19 5.3.2.2 Drywell plus Suppression 278,000 ft3 9.4, Item 19 Chamber Free Air Volume 5.3.2.3 Containment Elemental Standard Review Plan 6.5.2 9.4, Item 13 Iodine Removal Model 5.3.2.4 Drywell Surface Area for 32,250 R2 9.4, Item 13 Deposition/Plateout Model 5.3.2.5 Particulate (Aerosol) Powers' 10 percentile model 9.4, Item 14 Deposition/Plateout Model 5.3.2.6 Reactor Building 4,500,000 ftW. 9.4, Item 22 (Secondary Containment) Free Volume 5.3.2.7 Containment Leak Rate 3.0 v/1o/day (O to 30 days) Assumed into Reactor Building 5.3.2.8 Fraction of Containment 0.0 9.4, Item 24 Leakage that Bypasses the Standby Gas Treatment System (SBGTS) due to High Winds 5.3.2.9 Fraction of Reactor 0.5 . 9.1, RGP A.4.4 Building Available for Mixing 9.4, Item 23 5.3.2.10 SBGTS Exhaust Ratc 4,000 cfmi 10% 9.4, Item 21 CC-AA-309-1001, Rev 2

CALCULATION NO. DRE05-0048 REVISION NO. 0 PAGE NO. 22 of 64 Design Input Parameter Value Assigned Reference 5.3.2.11 SBGTS Exhaust Charcoal and HEPA Filter Efficiencies Elemental Iodine 50% Assumed Organic Iodide 50%

Particulate Aerosols 99%

5.4 ESF Leakage Model Parameters' 5.4.1 Suppression Pool Water 110,000 ft3 l 9.4, Item 27 Volume I I 5.4.2 Sump Water Activity (Ref. 9.1, RGP A.S.1, A.5.3 & Tables I & 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 5.4.3 ESF Leakage Rate 2 galmin (= 2 x 1 gal/min Assumed; 9.1, RGP A5.2 allowable leakage rate) 5.4.4 ESF Leakage Initiation 0 to 30 days 9.4, Item 31 Time and Duration 5.4.5 Suppression Pool not credited 9.1, RGP A.3.5 Scrubbing 5.4.6 Long-Term Suppression 7.53 9.12, page 11, 9.1, RGP A.2 Pool Water pH 5.4.7 Fraction of Iodine in ESF 0.10 9.4, Item 29; 9.1, RGP A.5.5 Leakage that becomes Airborne 5.4.8 Chemical Form of Iodine in ESF Leakage Elemental 97% 9.1, RGP A.5.6 Organic 3%

5.4.9 Fraction of Reactor 0.5 9.4, Item 32 Building Available for ESF Leakage Mixing 5.4.10 Percentage of ESF 100% 9.4, Item 33 Leakage that is filtered by the SBGTS 5.5 MSIV Leakage Model Parameters 5.5.1 Total MSIV Leak Rate 150 scfh @ 48 psig for 0 to 30 Assumed Through All Four Lines days 5.5.2 MSIV Leak Rate Through 60 seth @ 48 psig for 0 to 30 days Assumed - maximum leakage rate One Line With MSIV Failed through any one line 5.5.3 MSIV Leak Rate Through Three Intact Lines First Intact Line 60 scfh @ 48 psig for 0 to 30 days Assumed - maximum leakage rate through any one line Second Intact Line 30 scfh @ 48 psig for 0 to 30 days Assumed - remainder of unallocated leakage Third Intact Line 0 scfh 48 psig for 0 to 30 days Assumed - remainder of unallocated leakage CC-AA-309-1 001, Rev 2

-I -CALCULATION-NO.PREOS-0048 _-. -- LREV'ISION NO. 0 I-- PAGE NO. 23 of 64 1 Design Input Parameter Value Assigned Reference 5.5.4 Natural Removal 50 percent 9.5, Appendix B, page B-3 Efficiency For Elemental Iodine In Each Steam Line Volume 5.6 Control Room Model Parameters 5.6.1 CR Envelope Pressure 81,000 ft3 9.4, Item 34 Boundary Free Volume 5.6.2 CREV Filtration System 40 minutes 9.4, Item 40 Actuation Time Following a LOCA 5.6.3 CR Normal Operation 2,000 cfm +/- 10% 9.4, Item 41 Unfiltered Ventilation Air Intake 5.6.3 CR Emergency Ventilation 2,000 cfmn 10% 9.4, Item 42 Mode Air Intake Rate 5.6.4 CR Emergency Ventilation 0 cfm 9.4, Item 45 Mode Air Recirculation Rate though Filters 5.6.5 CR Unfiltered Inleakage 2000 cfin (includes ingress/egress Assumed during Normal Operation inleakage of 10 cfin) 5.6.6 CR Unfiltered Inleakage 400 cfm (includes ingress/egress Assumed during Emergency Ventilation inleakage of 10 cfm)

Mode 5.6.7 CR Emergency Ventilation Mode Intake Charcoal and HEPA Filter Efficiencies Elemental Iodine 99% Section 7.10 Organic Iodide 99%

Particulate Aerosols 99%

5.6.8 CR x/Qs For Containment & ESF Leakage Releasc Via SBGTS Stack (Station Chimney) Release Time X/Q (secrn 3) 0-2 6.42E-06 9.11, Table 3-5 2-8 2.87E-06 8-24 1.92E-06 24-96 8.03E-07 96-720 2.29E-07

,5.6.9 CR X/Qs For MSIV Leakage Release Via Units 2 an 3 MSIV Time X/Q (sec/n)_

0-2 1.30E-03 9.11, Table 4-1 2-8 1.06E-03 8-24 4.49E-04 24-96 2.96E-04 96-720 2.44E-04 CC-AA.309-1001, Rev 2

_ -CALCULATaON-NO-DRE05-0048 --jREISION NO. 0 -I-PAGE-NO of 64 I Design Input Parameter Value Assigned Reference 5.6.10 CR Occupancy Factors Time (Hr)  %

0-24 100 9.1, RGP 4.2.6 24-96 60 96-720 40

-5.-6.1 1CR-Breathing Rate -- 3.5E-04rm /scc 9.1, RGP 4.2.6 5.7 Offsite Dose Receptor Release Model Parameters 15.7.1 EAB X/Qs For Containment & ESF Leakage Release Via SBGTS Stack (Station Chimney) Release Time (hrs) X/Q (sec/m 3) 0-0.5 8.74E-05 9.1 1, Table 4-1 0.5-720 6.74E-06 5.7.2 EAB X/Q For MSIV Leakage Release Time (hrs) X/Q (sec/m 3) 0-720 2.51E-04 9.1 1, Table 4-1 5.7.3 EAB Breathing Rate 3.5E-04 m3/sec 9.1, RGP 4.1.3 5.7.4 LPZ X/Qs For Containment & ESF Leakage Release Via SBGTS Stack (Station Chimney) Release Time (hrs) XIQ (sec/mr) 0-0.5 8.84E-06 9.11, Table 4-1 0.5-2 1.78E-06 2-8 8.50E-07 8-24 5.87E-07 24-96 2.63E-07 96-720 8.3 lE-08 5.7.5 LPZ X/Qs For MSIV Leakage Release Time (hrs) X/Q (sec/rn 3) 0-2 2.63E-05 9.11, Table 4-1 2-8 1.09E-05 8-24 7.02E-06 24-96 2.70E-06 96-720 6.86E-07 5.7.6 LPZ Breathing Rates Time (hrs) BR (m 3/sec) 0-8 3.5E-04 9.1, RGPs 4.1.3 & 4.4 8-24 1.8E-04 24-720 2.31E-04 CC-AA-309-1001, Rev 2

CALCULATION NO. DREO5-0048 REVISION NO. 0 PAGE NO. 25 of 64 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:

RADTRAD (Ref. 9.2): This is an NRC-sponsored code approved for use in determining control room and offsite doses from releases due to reactor accidents. It was used in Dresden Calculation DREO1-0040, Revision 0 and Quad Cities Calculation QDC-0000-N-1 117, Revision

0. Among other things, it determined Reactor Building activity as a function of time. This output was used to develop the source strength for input into the MicroShield code.

MicroShield (Ref. 9.26): A commercially available and accepted code used to determine dose rates at various source-receptor combinations. Several runs were made at various times during the LOCA since the source strength varies over time.

6.2 Compliance With Reiulatorv 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 2

SALCULATIONAO-DRE0S0048 .- lEVISION NO. 0 _-[PAGENO.-26 of 64 7.0 CALCULATIONS 7.1 Dresden Plant Specific Nuclide Inventory File (NIF) For RADTRAD3.03 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, Dresden nuclide inventory file DPS-deftxt is compiled based on the fission products in the reactor core obtained from Reference 9.6.

7.2 Determination of MSIV Leak Rates 7.2.1 Proposed Case The total Icakage from all main steam lines is proposed to increase from 79.6 scfh to 150 scfh measured at 48 psig, allowing a maximum of 60 scfh @ 48 psig from any one of the 4 main steam lines.

The total containment leakage is also proposed to increase from I 0/o/day to 3 %/o/day,which includes leakage through the MSIVs.

7.2.2 MSIV Leakage During 0-2 hrs Drywell volume = 1.58E+05 ft3 (Ref. 9.4, Item 19)

Total MSIV leakage measured @ 48 psig = 150 scfh (assumed)

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 = 43.9 psig (Ref. 9.10, Table 4-1)

T @LOCA = Drywell peak temperature = 2910 F (Ref. 9.10, Table 4-1) = 291OF + 460 = 751OR P @STP = Standard pressure = 14.7 psia T @STP = Standard temperature = 680 F = 680 F + 460 = 5280 R V @STP = MSIV leakage based @ 48 psig = 150 scfh V @LOCA = (PV/T @STP) x (T/P @LOCA) 0-2 hrs MSIV leakage @ drywell peak pressure of 43.9 and temperature of 291OF

= 150 scfh x [14.7 psia / (43.9 psig + 14.7 psia)] x [751OR / 528t R]

= 150 scfh x 0.2509 x 1.422 = 53.52 cfh

= (53.52 ft3/hr x 24 hr/day) x 100% / 1.58E+05 ft3 0.813 0/olday 3

= (53.52 ft /hr) cfh / (60 min/hr) = 0.892 cfm 0-2 hrs Total proposed containment leakage = 3 %/day (Ref. 9.4, Item 18b) 0-2 hrs containment leakage into Reactor Bldg = 3 0/o/day - 0.813 0/o/day 2.187 D/o/day This containment leakage released into the Reactor Building is exhausted to the environment via the Standby Gas Treatment System (SBGTS).

The 0-2 hrs 150 scth MSIV leakage is released via three of the four Main Steam (MS) lines. A maximum allowable leak rate of 60 scfh is postulated from MS Line I with its failed MSIV. A I CC-AA-309-1001, Rev 2

_CALCULATION NO. DREO5-0048 REVISION NO. 0 PAGE NO. 27 of 64 maximum allowable leak rate of 60 scfh is postulated from intact MS Line 2. The remainder of 30 scfh is postulated from intact MS Line 3. No leakage is postulated from intact MS Line 4.

0-2 hrs allowable leakage from MS Line I with failed MSIV (at maximum 60 scfh leak rate)

= (60 scfh / 150 scfh total) x 53.52 cfh = 21.41 cth = 0.357 cfm 0-2 hrs allowable leakage from intact MS Line 2 (at maximum 60 scfh leak rate)

= (60 scfh / 150 scfh total) x 53.52 cfh = 21.41 cfb = 0.357 c&f 0-2-hrs-allowable-leakage-from intact-MS-Line -3(at balance of 30 scffi leak rate)-- -

= (30 scfh / 150 scfh total) x 53.52 cfh = 10.70 cfh = 0.178 cfm 7.2.3 MSIV Leakage During 2-720 hrs Two hours after a LOCA the drywell and suppression chamber volumes are expected to reach an equilibrium condition and the post-LOCA activity is expected to be homogeneously distributed between these volumes. The homogeneous mixing in the primary containment will decrease the activity concentration and therefore decrease the activity release rate through the MSIVs. To model the effect of this mixing, the MSIV flow rate used in the RADTRAD model is decreased by calculating a new leak rate based on the combined volumes of the drywell and suppression chamber.

Drywell + Suppression Chamber free air volume = 2.78E+05 fW (Ref 9.4, Item 19) 2-720 hrs MSIV leakage @ dxywell peak pressure of 43.9 psig = 53.52 cfh (per above)

=(53.52 cfh x 24 hr/day) x 100%/ 2.78E+05 ft3 = 462 %/oday 2-720 hrs Total proposed containment leakage = 3 %fday 2-720 hrs containment leakage into Reactor Bldg = 3 %/day - 0.462 %/day = L2/538 oday Corresponding MSIV leak rate = 53.52 cfh x (1.58E+05 ft3 / 2.78E+05 ft3) = 30.42 cfh 2-720 hrs allowable leakage from MS Line I with failed MSIV

= (60 scfh / 150 scfh total) x 30.42 cfh = 12.17 cfh = 0.203 cfm 2-720 hrs allowable leakage from intact MS Line 2

= (60 scfh / 150 scfh total) x 30.42 cfh = 12.17 cfh = 0.203 cfm 2-720 hrs allowable leakage from intact MS Line 3

= (30 scfh / 150 sefh total) x 30.42 cfi = 6.08 cfh = 0.101 cfm 7.2.4 MSIV Leakage To Environment It is assumed that the post-LOCA activity released in the SL with the failed inboard MSIV is instantaneously and homogeneously distributed in the single volume of SL between the RPV nozzle and outboard MSIV (well mixed volume). The MSIV leakage from the outboard MSIV expands to the atmospheric condition as follows:

Upstream of outboard MSIV (Section 7.2.2):

VI = 21.41 cfh PI = 43.9 psig + 14.7 = 58.6 psia Tl = (291 0F + 460) = 7510 R Dowvnstream of outboard MSIV (Atmospheric Condition):

V2 = TBD P2 = 14.7 psia T2 = (680 F + 460) = 5280 R CC-AA-309-1001, Rev 2

l CALCULATION NO. DRE050048 l REVISION NO.0 PAGE NO. 28 of 64 MSIV Leakage to Environment From MSIV Failed Line (MS Line 1):

V2 = (PV/T @1) x (TIP @2)

= (58.6 psia x 21.41 cfh / 751 OR) x (528OR / 14.7 psia)

= 60 cfh = 1.00 cfm This is as expected, given that the 21.41 cfh leakage rate is equivalent to 60 scfh upstream of the outboard MSIV, and therefore it is equivalent to 60 cfh downstream of the outboard MSIV in the presence of standard pressure and temperature atmospheric conditions.

The steam trapped between the MSIVs in the other two intact lines at the onset of a LOCA will at 1000 psia and 550°F (Rcf. 9.20, Item 9.16). The SL is insulated with 3-1/2" thick insulation (Ref. 9.20, Item 9.16). The steam line spools between the MSIVs will be at a considerably higher pressure (1000 psia - 58.6 psia = 941.4 psia) than the steam upstream of the inboard MSIV and the atmosphere downstream of the outboard MSIV. This extremely high positive pressure gradient across the MSIVs will prevent the MSIV leakage from migrating through the pipe spool between the MSIVs. To the contrary, the steam content in the pipe spool will leak until a negative pressure gradient is established across the inboard MSIV due to condensation of the steam in the spool. The time to establish the negative pressure gradient is considerably long. Therefore, to promote the MSIV leakage, it is conservatively assumed that the steam in the spool immediately cools down to atmospheric conditions, thereby establishing a negative pressure gradient across the intact inboard MSIV.

Upstream of inboard MSIV in intact MS Line 2 (Section 7.2.2):

VI = 21.41 cfh PI = 43.9 psig + 14.7 = 58.6 psia TI = (291°F + 460) =7510 R Downstream of inboard MSIV (assumed Atmospheric Condition):

V2 = TBD P2 = 14.7 psia T2 = (680 F + 460) = 528OR MSIV Leakage to Pipe Spool Between MS Line 2 MSIVs:

.V2 = (PVIT 1) x (T/P@2)

= (58.6 psia x 21.41 cfh / 751 0 R) x (528 0 R / 14.7 psia)

= 60 cfh = 1.00 cfin Upstream of outboard MSIV (i.e., downstream of inboard MSIV) in intact MS Line 2:

V2 = 60 c&h P1 = 14.7 psia T2 = (680 F + 460) = 5280 R Downstream of outboard MSIV in intact MS Line 2 (assumed Atmospheric Condition):

V3 = TBD P2 = 14.7 psia T2 = (680 F + 460) = 528 R MSIV Leakage to Environment From MS Line 2:

V3 = (PVrT@2) x (T/P@3)

= (14.7 psia x 60 cfh / 528°R) x (528 0 R1 14.7 psia)

= 60 cfh = 1.00 cfm This is as expected, given that the pressure and temperature conditions in the pipe spool between the MS Line 2 MSIVs are assumed to be the same as the standard pressure and temperature atmospheric conditions present in the environment.

CC-AA-309-1001,Rev2

_lCALCULATION-NO.DRE05-0048 4 ..REYISIONNO.0 j-PAGE-NO. 29 of 64 A similar calculation using the same pressure and temperature conditions results in the MSIV Leakage of 30 cfh (0.50 cfm) into the pipe spool between the MS Line 3 MSIVs, and from the pipe spool to the Environment.

The 2-720 hr MSIV leakages to Environment Per Section 7.2.3, two hours after a LOCA the d-ywell and suppression chamber volumes are expected

.toreach-anequilibriumcri -dition-and-th -post-LOCAac-tivity isex-pe-ctedto'be h6fdgelneously distributed between these volumes. Therefore, the leak rates based on the activity in the drywell are not applicable during this period. This results in a reduction in the 0-2 hr MSIV leakages to the environment by the ratio of the drywell volume to the combined drywell plus suppression volume:

2-720 hrs MSIV leakage release to environment from MS Line 1 with failed MSIV

= 60 cfh x (1.58E+05 f 3 / 2.78E+05 ft3) 34.10 cfh = 0.568 cfm 0-720 hrs MSIV leakage release to environment from intact MS Line 2

= 60 cfh x (1.58E+05 ft3 / 2.69E+05 ft 3) = 34.10 cfh = 0.568 cfm 0-720 hrs MSIV leakage release to environment from intact MS Line 3

=30 cfh x (1.58E+05 ft3 / 2.69E+05 ft3) = 17.05 cfh = 0.284 cfm 7.3 Main Steam Line Volumes & Surrace Area For Plateout of Activit A comparison of the total horizontal lengths between the reactor pressure vessel and the outboard MSIV of the three shortest main steam piping runs of QCNPS, DNPS Unit 2 and DNPS Unit 3 is shown in Table 1. It can be seen from the comparison that the QCNPS main steam piping runs are shorter, and would therefore result in less (i.e., conservative) aerosol deposition. Therefore, the main steam line lengths, areas, and volumes for plateout of activity calculated for the QCNPS design in Reference 9.20 are applicable to the Dresden design. The following summarizes these dimensions that are used in Tables 2 and 4.

7.3.1 Piping Line 2-3001A-20" from RPV Nozzle N3A to Outboard Isolation Valve with MSIV failed (70 scfi)

Control Volume V1 for MSIV Failed SL Between RPV Nozzle & inboard MSTV (60 scfh)

Total Volume Vii = 152.96 f 3 Horizontal pipe volume VH11 = 40.00 ft Horizontal pipe length for gravitational aerosol deposition LHII = 23.42' Horizontal pipe projected surface area for gravitational aerosol deposition AH11 = 34.54 ft2 CC-AA.309-1001, Rev2

CALCULATION NO. DRE05-0048 I REVISION NO. 0 PAGE NO.'30 of 64 Control Volume V1I for MSIV Failed Line Between Inboard & Outboard MSIVs (60 scfh)

Total volume V 12 = 47.28 ft3 Horizontal pipe volume VH 1 2 = 47.28 ft Horizontal pipe length for gravitational aerosol deposition 1,12 = 27.68' Horizontal pipe projected surface area for gravitational aerosol deposition A1112 = 40.83 ft2 Control Volumes VI1 + V1 t for MSIV Failed SL Between RPV Nozzle & outboard MSIV (60 scfl)

Total Volume VI = VI1 + V1 2 = 152.96 ft3 + 47.28 ft3 = 200.24 ft 3 (Used in RADTRAD Runs DRE400MS3I.psf and DRE400MS32.psf)

Total Horizontal pipe volume VHI = VII1 + VH12 = 40.00 ft3 + 47.28 ft3 = 87.28 ft3 (Used in Table 2)

Total Horizontal pipe length for gravitational aerosol deposition LI = L1il, + LH 1 2 = 23.42' + 27.68' = 51.10' Total Horizontal Surface Area AII = A 111 + AH1 2 = 34.54 ft2 + 40.83 ft2 = 75.37 ft2 (Used in Table 2) 7.3.2 First Intact SL 2-3001D-20" from RPV Nozzle N3D to Outboard MSIV (60 scfh)

Control Volume 2 for First Intact SL Between RPV Nozzle & Inboard MSIV (60 scfh)

Total volume V 2 =152.93 ft3 Horizontal pipe volume VH2 = 39.97 ft3 Horizontal pipe length for gravitational aerosol deposition LH2 = 23.40' Horizontal pipe projected surface area for gravitational aerosol deposition A112 = 34.52 ft2 Control Volume 3 for First Intact SL Between Inboard & Outboard MSIVs (60 scfh)

Total volume for first intact pipe between Inboard & Outboard MSIVs CC-AA-309-1 001, Rev 2

_CALCULATION NO. DRE05-0048 __j REVISION NO. 0 -_

- PAGE NO. 31 of -64 V3 = 49.11 ft3 Horizontal pipe volume VH3=49.11 ft Horizontal pipe length for gravitational aerosol deposition Lto = 28.75' Horizontal pipe projected surface area for gravitational aerosol deposition AH3 = 42.41 ft2 7.3.3 Second Intact SL 2-3001C-20" from RPV Nozzle N3C to Outboard MSIV (30 scfh)

Control Volume 4 for Second Intact SL Between RPV Nozzle & Inboard MSIV (30 scfh)

Total volume V 4 = 163.75 ft3 Horizontal pipe volume VH4 = 49.01 ft3 Horizontal pipe length for gravitational aerosol deposition LJR4 = 28.70' Horizontal pipe projected surface area for gravitational aerosol deposition A1 4 = 42.33 ft2 Control Volume 5 for First Intact SL Between Inboard & Outboard MS1Vs (30 scfh)

Total volume for first intact pipe between Inboard & Outboard MSIVs V5 = 49.11 ft3 Horizontal pipe volume V,,5 = 49.11 ft3 Horizontal pipe length for gravitational aerosol deposition L = 28.75' Horizontal pipe projected surface area for gravitational aerosol deposition AHS = 42.41 ft2 7.4. Aerosol Deposition On Horizontal Pipe Surface The DNPS main steam piping from the reactor pressure vessel (RPV) nozzle to the outboard MSIV is ASME Class I seismically analyzed to assure the piping wall integrity during and after a seismic (safe shutdown earthquake [SSE]) event. RG 1.183, Appendix A, Section 6.5 requires that the components and piping systems used in the release path are capable of performing their safety function during and following a SSE. The main steam lines credited in the MSIV leakage path are qualified to withstand the SSE, therefore, these lines are credited for the aerosol deposition in the following section:

CC.AA-309-1001, Rev 2

I CALCULATION NO. DRE05-0048 IREVISION NO. 0 l PAGE NO. 32 of 64 The Brockmann model for aerosol deposition (Ref. 9.2, 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 9.5, 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. 9.5). Therefore, the following aerosol deposition model is used, which is accepted by the Staff in Reference 9.5, Appendix A). Therefore, the Staff performed a Monte Carlo analysis to determine the distribution of aerosol settling velocities for the main steam line during the in-vessel release phase. The accepted 40 percentile settling velocity is reasonably conservative for aerosol deposition in the MSIV leakage. The results of the Monte Carlo analysis for settling velocity in the main steam line are given in the following Table:

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

60t (average) 0.00148 11.43 50th (median) 0.00117 9.04 40th 0.00081 6.26 10th 0.00021 1.62 7.4.1 MSIV Failed Line The derivation of staff's well-mixed model begins with a mass balance as follows (Ref. 9.5, Page A-2):

V

• dC = Q

• Cin - Q

• C- X.

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

Where us = settling velocity A = settling area The aerosol settling velocities in the different control volumes are calculated in Table 2 using the above equation based on the horizontal pipe projected areas and well mixed horizontal volumes obtained from Section 7.3.

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

CaCm* 1 1+ RE*V Q

l. CC-AA-309-1001, Rev 2

CALCULATION NO. DREOS-0048 RqEVtISION NO. 0 . -PAGENO. 33 of 64l 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:

fi 1- C = - 1 (2)

Cin e1+ x1*V Q

Equation (2) is used to calculate the aerosol removal efficiencies in Table 4. Note that the volumetric flow rate used to determine the removal efficiency is the full flow rate through the line (60 or 30 cfm).

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

I gallon/min x 2 x 1/7.4805 fQ/gallon = 0.2674 cfm 10% of ESF leakage becomes airborne = 0.10 x 0.2674 = 0.02674 cfm 7.6 External Cloud Gamma Dose Attenuation Factor The gamma attenuation for concrete shielding for an external cloud dose is conservatively calculated for an average gamma energy of 1.0 Mev.

The gamma radiation external radioactive plume shine to the CR personnel is attenuated by the 1'-6" minimum concrete wall shielding (Ref. 9.22.c). Gamma dose attenuation for 1'-6" concrete shielding is calculated as follows:

Mass attenuation coefficient for concrete at I Mev p/p = 0.0635 cm 2 /g (Ref 9.14, Table 3.7)

Density of concrete p = 2.3 g/cm3 (Ref. 9.14; Table 11.3)

Linear attenuation coefficient 1p in concrete'= p/p x p = 0.0635 cm2 /g x 2.3 g/cm 3 = 0.146 cmf1 Shielding thickness r = 18 inch x 2.54 cm/inch = 45.72 cm pr in concrete shielding = 0.146 cm~1 x 45.72 cm = 6.675 mean free paths Exposure buildup factor for isotropic point source at disintegration energy of I Mev and 6.675 mean free paths of the I Mcv gammas Bp(pr) =Al e 1'1 +A2 eC2 1" (Ref. 9.14, page 428)

Where Al, A2 , ab,and a2 are functions of energy, and Ai + A2 = I Values of these parameters are obtained from Table 10.3 of Reference 9.14 for 1 Mev gamma in concrete shielding as follows:

Al = 25.507 -a, = 0.07230 a2 =-0.01843 A 2 = 1 -Al = 1-25.507=-24.507 [tr=6.675 Substituting these values in the above equation yields:

B,(pir) = 41.32-27.71 = 13.61 Direct Shield Attenuation Io = Bp(gir) eM' Where CC-AA-309-1001, Rev2

__CALCULATION NO. DRE05-0048 _ REVISION NO. 0 -PAGE NO. 34 of 64 I = shielded gamma dose rate lo = unshielded gamma dose rate Bp(pr) = Exposure buildup factor Substituting the values of parameters into the above attenuation Equation (I) yields a direct shield attenuation factor of

-- -Bar) -e-163.61

~r-=16 -667)13.-61361 x1.262E 0.0172 7.7 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 (page 6.5.2-

10) (Ref. 9.9) as follows:

The equation for the wall deposition is:

K x .AJV Where:

= first order removal coefficient by wall deposition Kw= mass transfer coefficient = 4.9 m/hr (Ref. 9.9, page 6.5.2-10)

A = wetted surface area = 32,250 ft2 (Ref. 9.4, Item 13)

V = drywell net free air volume = 1.58E+05 fl3 (Ref. 9.4, Item 13) w=x A/V = 4.9 rn/hr x (3.2808 ft/m) (32,250 ft2) / (1.58E+05 ft3) = 3.28 hf' Maximum DF of elemental iodine 200 I/DF=e e I /200=e (-3.28t) 0.005 = e (-3.28t)

In (0.005) = -3.28t

-5.298 = -3.28 t t= 1.615 hr The maximum iodine activity concentration takes place in the containment at the end of the early-in-vessel release phase (Ref. 9.1, Appendix A, Section 3.3), which is at 2.0 hr after the onset of a LOCA (Ref. 9.1, Tables I & 4).

Termination time for elemental iodine removal by wall surface deposition

=2.0hr+ 1.615 hr 3.615 hr 7.8 Containment Shine Shielding Geometry Reactor Building Shielding Parameters: (Ref 9.21)

CC-AA-309-l1OO, Rev 2

l CALCULATION.NO.DREOS-0048, . . _ REYJSIONANO.0 -<PAGE 6 NO.35of Length = 147'-0" Width= 117'-6" Height = 659-6" - 613'-0" - 2'-0" (for roof thickness) = 44'-6" z 44'-0" used in the analysis to adjust the roof thickness dimension Volume of Source = 147' x 117.5' x 44' 759,990 ft3 ( 2.15E+10 cm3 ) used in the analysis Distance between north-east corner of RB and normally occupied CR area = Distance between Columns 32 through & 38 20'-7" + 20'-7' + 20'-7" + 25'-0" + 22'-9" + 21 '-6-1/2" = 131 '-0-1/2" (Ref. 9.21.a)

Elevation difference between CR operator and RB operating floor

= 613'-0" RB operating floor elevation - [(534'-0") CR floor elevation + 7'-O" height of operator (assumed)]

= 613' - 541'-0' = 72' Line of sight distance between CR operator location and centerline of RB source

= [(72 )2 + (131.04t)2]t2 = 150' To this line of sight distance I '-0" is added representing the air separation between Column 32 and the CR operator, and 2'-6" is subtracted representing concrete shielding (per Section 2.4.3). The result is an effective line of sight air attenuation distance of 148'-6" (see Figures 4 & 5).

Gamma dose rate reduction factor based on RB volume = 759,990 & / 2.35E+06 ft3 = 0.3234 7.9 CR Containment Shine Dose Per Section 2.4.3, the shielding configuration (line of sight distance and intercepting concrete shielding) for the Dresden CR operator dose are virtually identical to Quad Cities. For the given shielding geometry, the CR dose is dependent on the RB source term. The comparison of the total isotopic activities in Table 7 with the corresponding isotopic activities in the Quad Cities RB in Reference 9.20, Table 6 indicates that the isotopic activities in the Quad Cities RB arc slightly higher than those in the Dresden RB. Therefore, the post-LOCA containment shine dose of 0.22 rem (Ref. 9.20, Sections 7.8 &

7.9) calculated for the Quad Cities CR operator can be conservatively applied to the Dresden CR operator. The resulting containment shine dose is listed in Section 8.1.

7.10 SGTS Vent and CR Charcoal Filters Efficiencies Technical Specification 5.5.7, Ventilation Filter Testing Program (VFTP), requires routine testing of safety related filtration systems.

In-place penetration and system bypass testing on HEPA filters is routinely performed in accordance with RG 1.52 Revision 2 and ANSI/ASME N510-1980. The acceptance criteria for the safety related HEPA filters are as follows:

CC-AA-309-1001, Rev 2

CALCULATION NO. DREO5-0048 I REVISION NO. 0 PAGE NO. 36 or 64 SGTS: <1%

CREV: <0.05%

In-place penetration and system bypass testing on charcoal adsorbers is routinely performed in accordance with RG 1.52 Revision 2 and ANSI/ASME N510-1980. The acceptance criteria for penetration and system bypass testing of the safety related charcoal adsorbers are as follows:

SGTS: <1%

CREV: <0.05%

• Laboratory testing of charcoal samples are obtained in accordance with RG 1.52 Revision 2 and tested in accordance with ASTM D3803-1989. The acceptance criteria for the safety related charcoal adsorbers are as follows:

SGTS: 2.5%

CREV: 0.5%

If all of the above requirements are met (assuming a safety factor of two for the laboratory testing) the Regulatory Guide 1.52 assigned efficiencies may be applied to the charcoal adsorbers (99% for HEPA filters). For Dresden, these are:

SGTS: 95%

CREV: 99%

7.11 Post-LOCA CREV Filter Shine Dose The post-LOCA CREV filter shine dose due to the MSIV leakage is calculated in the following sections.

The containment and ESF leakages contribute insignificant CR dose (Section 8.1). Therefore, they are not considered in the filter shine dose analysis.

7.11.1 Iodine Deposition on CREV Charcoal Filter-MSIV Leakaae Tables 9 and II document the elemental iodine atoms and organic iodide atoms released to the environment from the three main steam lines modeled with MSIV leakage for time intervals of 0,6667 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, 2 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, 8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, 24 to 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br />, and 96 to 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> as determined in RADTRAD file DRE400MS3l.oO. These time intervals coincide with the varying atmospheric dispersion factor defining MSIV leakage releases to the CREV system intake louvers. There is no filter activity loading prior to the initiation of the CREV system at 40 minutes.

For each time interval, Tables 10 and 12 multiply the iodine atoms released to the environment, with the atmospheric dispersion factor, the CREV filtered intake flow, and the charcoal filter efficiency. The result is the total number of elemental and organic iodine atoms drawn into, and retained on, the CREV charcoal filter.

The combined total of elemental and organic iodine atoms retained on the CREV charcoal filter is:

= 1.425+15 elemental iodine atoms (Table 10) + 1.277E+16 organic iodide atoms (Table 12)

= 1.4195E+16 elemental + organic iodine atoms.

I CC-AA-309-1001, Rev 2

l CALCULATION NO. DRE05-0048 REVISION NO. 0 PAGE NO. 37 of 64 The iodine atom/curie relationship is established using the containment leakage run DRE400CL3 1.o0file as shown in Table 15, which is a typical relationship for all release paths.

The total (elemental + organic) iodine activity deposited on the CREV charcoal filter due to the MSIV leakage is calculated in Table 16 using this iodine atom/curie relationship and the combined total of elemental and organic iodine'atoms retained on the CREV charcoal filter. A review of Table 16 documents that the accumulation of un-decayed iodine activity on the CREV charcoal filter of approximately 2.2 curies is insignificant. This is as expected, because most of the elemental iodine is removed by elemental deposition in the main steam piping before it is released to the environment and it is further reduced by air dilution before it migrates to the CR air intake. The natural radioactive process will further decay the iodine on the CREV charcoal bed.

7.11.2 Aerosol Mass Deposited On CREV HEPA Filter - MSIV Leakane:

Table 13 documents the aerosol mass released to the environment from the three main steam lines modeled with MSIV leakage for time intervals of 0.6667 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, 2 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, 8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, 24 to 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br />, and 96 to 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> as determined in RADTRAD file DRE400MS3 I .oO. These time intervals coincide with the varying-atmospheric dispersion factor defining MSIV leakage releases to the CREV system intake louvers. There is no filter activity loading prior to the initiation of the CREVS at 40 minutes.

For each time interval, Table 14 multiplies the aerosol mass released to the environment, with the atmospheric dispersion factor, the CREV filtered intake flow, and the HEPA filter efficiency. The result is the total aerosol mass drawn into, and retained on, the CREV..HEPA filter is 9.233E-07 kg (Table 14).

The aerosol mass/curie relationship is established using the containment leakage run DRE400CL3 1.oO file as shown in Table 17, which is a typical relationship for all release paths.

The total aerosol activity deposited on the CREV HEPA filter due to the MSIV leakage is calculated in Table 18 using this aerosol mass/curie relationship and the total'aerosol mass retained on the CREV charcoal filter. A review of Table 18 documents that the accumulation of aerosol activity on the CREV HEPA filter (no isotope with more than le-4 curies) is insignificant. This is as expected, because most of aerosol deposit out in the main steam piping horizontal surface before being'released to the environment (see Table 4 for the aerosol removal efficacies due to gravitational deposition).

I CC-AA-309-1001, Rev 2

CALCULATION NO. DRE05.0048 I REVSION NO.0 PAGE NO.38 of 64 8.0 RESULTS

SUMMARY

& CONCLUSIONS 8.1 Results Summary The results of AST analyses for the proposed licensing basis are summarized in the following table:

Post-LOCA. Post-LOCA TEDE Dose (Rem)

Activity Release Receptor Location Path Control Room EAB LPZ Containment Leakage 1.59E-02 6.77E-02 7.53E-02

._ (occurs @4.1 hr)

ESF Leakage 3.24E-02 2.85E-02 5.79E-02

- (occurs @ 16 hr)

MSIV Leakage 4.17E+00 1.48E+00 3.68E-Ol (occurs @ 2.9 hr)

Containment Shine to CR 2.20E-OI O.OOE+OO .O.OOE+OO External Cloud Shine to 2.73E-O O.OOE+OO O.OOE+OO CR CR Filter Shine to CR Negligible O.OOE+OO O.OOE+OO Total 4.71E+00 1.58E+00 5.01E-01 Allowable TEDE Limit 5.OOE+00 2.50E+01 2.50E+01 RADTRAD Computer Run No.

Containment Leakage DRE400CL31 DRE400CL31 DRE400CL31 ESF Leakage DRE400ESF31 DRE400ESF31 DRE400ESF31 MSIV Leakage DRE400MS31 DRE400MS31 DRE400MS31 8.2 Conclusions The Section 8.1 results of this analysis, using conservative as-built design inputs and assumptions that reflect the proposed AST implementation indicate that the EAB, LPZ, and CR doses are within their allowable TEDE limits, I CC-AA-309-1001, Rev2

l CALCULATION NO. DRE05-0048 REVISION NO. 0 PAGE NO. 39 of 64

9.0 REFERENCES

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

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

9.3 I0CFR50.67, "Accident Source Term."

9.4 Exelon Transmittal of Design Information No. ER2002-9994, Rev 1, "Dresden Station Concurrence with the Design Inputs as established for Alternate Source Term (AST) LOCA Analysis," Revision 1, July 31, 2002.

9.5 AEB 98-03, Assessment of Radiological Consequences for the Perry Pilot Plant Application Using The Revised (NUREG-1465) Source Term.

9.6 GE Task Report No. GE-NE-A22-00103-64-01, Rev 0, Project Task Report: "Dresden and Quad Cities Asset Enhancement Program, Task T0802: Radiation Sources and Fission Products" Dated August 2000.

9.7 EPA-520/1-88-020, Federal Guidance Report 11, Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion and Ingestion."

9.8 EPA-402-R-93-081, Federal Guidance Report 12, "External Exposure to Radionuclides in Air, Water and Soil."

9.9 NUREG-0800, Standard Review Plan, "Containment Spray as a Fission Product Cleanup System," SRP 6.5.2, Revision 2, 1988.

9.10 GE-NE-A22-00103-08-01, Revision 1, Class 3, December 2000, Project Task Report T0400, Containment System Response 9.11 Exelon Calculation No. DRE04-0030, Rev 1, Atmospheric Dispersion Factors (X/Qs) for Accident Release.

9.12 Exelon Calculation No. DRE02-0033, Rev 0, Ultimate Suppression Pool pH Following a Loss of Coolant Accident.

9.13 USNRC, "Laboratory Testing of Nuclear-Grade Activated Charcoal," NRC Generic Letter 99-02, June 3, 1999 9.14 Introduction To Nuclear Engineering By John Lamarsh, Third Printing, December 1977, Addison-Wesley Publishing Company.

9.15 Dresden Unit 2 Field Isometric Drawings:

a. Unit 2 Field ISO No. 30-1, SHT 1, Rev 4, # 30 Main Steam

b. Unit 2 Field ISO No. 30-3, SHT 1, Rev 5, # 30 Main Steam

c. Unit 2 Field ISO No. 30-4, SHT 1, Rev 5, # 30 Main Steam

d. Unit 2 Field ISO No. 30-6, SHT 2, Rev 3, # 30 Main Steam

e. Unit 2 Field ISO No. 30-7, SHT 1, Rev 3, # 30 Main Steam

f. Unit 2 Field ISO No. 30-8, SHT 2, Rev 3, # 30 Main Steam 9.16 Dresden Unit 3 Field Isometric Drawings:

a. Unit 3 Field ISO No.30-300, SHT 1, Rev 2, # 30 Main Steam

b. Unit 3 Field ISO No.30-304, SHT 1, Rev 3, # 30 Main Steam CC-AA-309-1001, Rev2

l CALCULATION -OJDREOS0048 - IREVISION NO.0 -- 4PAGENO.40 of 64

c. Unit 3 Field ISO No.30-302, SHT 2, Rev 3, # 30 Main Steam

d. Unit 3 Field ISO No.30-306, SHT 1, Rev 3, # 30 Main Steam

e. Unit 3 Field ISO No.30-303, SHT 1, Rev 2, # 30 Main Steam

f. Unit 3 Field ISO No.30-307, SHT 1, Rev 3, # 30 Main Steam 9.17 Dresden Technical Specification No. 5.5.7, Ventilation Filter Testing Program, Paragraph c 9.18 Dresden UFSAR, Rev 4, Section 15.6.5, Loss of Coolant Accident Resulting from Piping Breaks Inside

-Containment - --- -

9.19 SWEC Calculation No. DRE01-0040, Rev 0, Site Boundary and Control Room Doses following a Loss of Coolant Accident using Alternative Source Terms.

9.20 Quad Cities Calculation No. QDC-0000-N-1481, Rev 0, Post-LOCA EAB, LPZ, & CR Doses Using Alternative Source Term 9.21 DPS General Arrangement Drawings:

a. M-3, Rev S, Mezzanine Floor Plan

b. M4, Rev AB, Ground Floor Plan

c. M-7, Rev D, Sections A-A & B-B

d. M-8, Rev C, Sections C-C & D-D

e. M-9, Rev H, Sections E-E & F-F 9.22 DPS Control Room Structure Drawings:

a. B-351, Rev H, Turbine Building Mezzanine Floor Plan, Control Room North Area

b. B-352, Rev E, Turbine Building Mezzanine Floor Plan, Control Room South Area

c. B-144, Rev AN, Turbine Building Framing Plan Elevation 534'-0" Control Room Area 9.23 DPS HVAC Drawings:

a. M-3123, Rev 2, General Arrangement Plan At EL. 534'-0", Control Room HVAC Upgrade

b. M-3124, Rev E, Control Room HVAC Equipment Rooms, Plans & Sections, EL. 534'-0" & EL.

549'-0"

c. M-3128, Rev 8, Piping Arrangement Control Room HVAC Area EL. 534'-0" & EL. 549'-0".

9.24 American Air Filter Drawing No. R107D-1327238-B, Housing Assembly Control Room Filter (Quad Cities Generating Station.

9.25 Standard Review Plan Section 15.0.1, Rev. 0, July 2000 9.26 MicroShield Computer Code, V&V Version 5.05, Grove Engineering CC-AA-309-lOO1 . Rev 2

ICALCULATION NO. DRE05-0048 I REVISION NO. 0 PAGE NO. 41 of 64 10.0 TABLES Table I Comparison of Horizontal Pipe Length for Aerosol Deposition (Measured between RPV and Outboard MSIV)

Horizontal Pipe Length (ft)

Quad Cities Dresden Dresden Steam Limiting Unit 2 Unit 3 Header Case ID A B C 213-3001A 51.10 53.69 53.69 2/3-3001-C 57.45 62.62 63.57 2/3-3001-D 52.15 53.70 53.69 A From Reference 9.20, Section 7.3 B From Reference 9.15 C From Reference 9.16 CC-AA-309-1001, Rev 2

_ CALCULATIONNO-DREOS-0048 _LREVISIONNO. 0 ---PAGE NO. 42 Or 64 I Table 2 Rate Constant (,) for Aerosol Settling In Main Steam Piping With MSIV Failure Intact Steam Line Without NISIV Failure RPV Nozzle A To Inboard MSIVI RPV Nozzle D Inboard MSIV2 RPV Nozzle C Inboard MSIV3 Parameter To To To To To Outboard MSIVI Inboard MSIV2 Outboard MSIV2 Inboard MISIV3 Outboard MSIV3 Control Volume Control Volume Control Volume Control Volume Control Volume VI. V2 V3 V4 VA_

Settling Velocity* 9564 (ft/hr) 9_54___56_95649 _54_956 Horizontal Settling 75.37 34.52 42.41 42.33 42.41 Horizontal Pipe 87.28 39.97 49.11 49.01 49.11 V olum e V I,, t)_ _ _ _ _ __ _ _ _ _ _ _ _ __ _ _ _ _ _ _ _

Rate Constant for Settling s, ( ) 8.259 8.260 8.259 8.260 8.259

• 40 Percentile Settling Velocity - 0.00081 m/sec (Ref. 9.5, Appendix A, Table A-1) x 3.28 fi/m x 3600 sec/hr = 9.564 ft/sec Main Steam Piping Parameters From Section 7.3 I CC-AA-309- 1001, Rev 2

l.ICALC~ULATJO9N 3HDDREOS OO048

.CALCJLATJONOJREO5AO48

- _ljJIEVISION.NO.

---RMBYSION.NO. O 0

--- 4 -PAGE-NO.A43 of 64 PAGENO.43 of 64 I Table 3 S1SIV Leak Rate In Different Control Volumes (150 scfh)

Sl'_Leak Rate From Drvwell To Main Steam Various Control Volumes (cfhy(cfm Post-LOCA Drywell To Volume VI Drywell To Intact Une I Volume V3 Drywell To Intact Line 2 Volume V5 Time MS1V Failed To Intact Line 1 Volume V2 To Intact Lne 2 Volume V4 To Interval Volume V] Atmosphere Volume V2 To Atmosphere Volume V4 To Atmosphere (hr) Volume V, Volume V _ _

0-2 21.41 60.00 21.41 60.00 60.00 10.70 30.00 30.00 0.357 1.000 0.357 1.000 1.000 0.178 0.500 0.500 2-720 12.17 34;10 12.17 34.10 34.10 6.08 17.05 17.05 0.203 0.568 0.203 0.568 0.568 0.101 0.284 0.284 MSIV Leak Rate Information From Section 7.2 I CC-AA-309-1001 Rev 2 I

CALCULATION NO. DREO5-0048 REVISION NO. 0 1 PAGE NO. 44 of 64 Table 4 Aerosol Removal Efficiency Due To Gravitational Deposition On Horizontal Pipe Surface Volume V, 152.96 ft2 Aerosol Volume V, - 49.11 fe Aerosol Post-LOCA Settling Horizontal Volumetric Removal Post-LOCA Settling Horizontal Volumetric Removal Time Rate Pipe Flow Efficiency Time Rate Pipe Flow Efficiency Interval Constant Volume Rate Interval Constant Volume Rate (br) (hr') (f1) (ft'lhr) (%) (1 (hr) (f') (ft'/hr) (%)

0-720 8.259 ' 40 60.00 84.63 0-720 8.259 49.11 60.00 87.11 Volume V12 47.28 f1 Aerosol Volume V.4 - 163.75 f1. Aerosol Post-LOCA Seling Horizontal Volumetric Removal Post-LOCA Settling Horizontal Volumetric Removal Time Rate Pipe Flow Efficiency Time Rate Pipe Flow Efficiency Interval Constant Volume Rate Interval Constant Volume Rate (hr) (hr ef') (ft'/hr) (% (hr) (hrl) (ft') (ft'llr) (%

0-720 8.259 47.28 60.00 86.68 0-720 8.260 49.01 30.00 93.10 Volume V2 - 152.93 fe Aerosol Vol me Vs - 49.11 f1 Aerosol Post-LOCA Settling Horizontal Volumetric Removal Post-LOCA Settling Horizontal Volumetric Removal Time Rate Pipe Flow Efficiency Time Rate Pipe Flow Efficiency Interval Constant Volume Rate Interval Constant Volume Rate

_2 (h8)6 6 8 0-720 (h3 _ .(he2 49.1 30.00') 93.1 070 8.260 39.97 60.00 84.62 0-720 8.259 49.11 30.00 931 MSIV Failed Line Well Mixed Volume V, Vl + V12 =152.96 t +47.28 1' =200.24 Used In RADTRAD Model (Section 7.3.1)

Note: The control volumes VI1 & V12 are combined in the analysis to postulate the failure of inboard MSIV to close. However, the aerosol and elemental iodine depositions in the pipe volume between RPV nozzle and inboard MSIV (Control Volume V,,) are not credited in the analysis by only using the removal efficiencies in Control Volume V12 .

I CC-AA-309-1001. Rev 2 l

_ CALCULATION NO. DREO5-0048 _ _ VISION NO. P

.--IY-A-GIENO.45-of 64 I Table 5 Post-LOCA Reactor Building Isotopic Inventory - Containment Leakage Post-LOCA Reactor Building Isotopic Inventory (Ci)

Isotope ' Containment Leakaaze 0.667 hr 2.0 hr 4.0 hrs 8.0 hrs 16 hrs 24 hrs Co-58 9.420E-03 6.357E-01 1.261 E+00 9.156E-01 3.663 E-0 I 1.430E-0I

-Co-60 -1.128E -7.614E201- -1.511E+00 - 1.099E+00. -4.412E-01 -1.728E-0I.

iKr-85 3.474E+01 9.007E+02 3.188E+03 6.410E+03 9.628E+03 1.082E+04 Kr-85m 4.862E+02 1.026E+04 2.664E+04 2.885E+04 1.257E+04 4.0952+03 Kr-87 7.145E+02 8.957E+03 1.066E+04 2.422E+03 4.645E+01 6.6652-01 Kr-88 1.228E+03 2.299E+04 4.995E+04 3.783E+04 8.064E+03 1.286E+03 Rb-86 3.773E+00 3.409E+01 5.944E+01 4.243E+01 1.679E+01 6.4942+00 Sr-89 1.197E+01 8.072E+02 1.601E+03 1.1622+03 4.641 E+02 1.81 OE+02 Sr-90 1.739E+00 1.174E+02 2.3312+02 1.696E+02 6.805E+01 2.666E+01 Sr-91 1.447E+01 8.863E+02 1.520E+03 8.261 E+02 1.8492+02 4.041E+01 Sr-92 1.398E+01 6.710E+02 7.986E+02 2.088E+02 1.083E+01 5.4822-01 Y-90 2.0022-02 2.426E+00 9.3022+00 1.365E+01 1.068E+01 6.062E+00 Y-91 1.5592-01 1.0682+01 2.1792+01 1.661E+01 7.056E+00 2.844E+00 Y-92 4.794E-01 1.416E+02 4.948E+02 3.502E+02 5.202E+01 5.473E+00 Y-93 1.857E-01 1.144E+01 1.9792+01 1.094E+01 2.536E+00 5.737E-01 Zr-95 2.190E-01 1.478E+01 2.9302+01 2.1282+01 8.5092+00 3.321 E+00 Zr-97 2.157E-01 1.379E+01 2.522E+01 1.557E+01 4.5002+00 1.270+400 Nb-95 2.201E-01 1.486E+01 2.9492+01 2.145E+401 8.609E+00 3.372E+00 Mo-99 3.134E+00 2.086E+02 4.0562+02 2.829E+02 1.044E+02 3.759E+01 Tc-99m 2.764E+00 1.864E+02 3.684E+02 2.633E+02 1.012E+02 3.764E+01 Ru-103 2.656E+00 1.791E+02 3.550E+02 2.575E+02 1.027E+02 4.001 E+O I Ru-lOS 1.6852+00 9.238E+01 1.3422+02 5.228E+01 6.018E+00 6.761E-01 Ru-106 1.132E+00 7.642E+01 1.5172+02 1.103E+02 4.424E+01 1.732E+01 Rh-105 1.776E+00 1.195E+02 2.341E+02 1.629E+02 5.761 E+O I 1.948E+01 Sb-127 3.678E+00 2.458E+02 4.807E+02 3.394E+02 1.283E+02 4.732E+01 Sb-129 9.831 E+00 5.359E+02 7.7182+02 2.955E+02 3.286E+01 3.5662+00 Te-127 3.672E+00 2.468E+02 4.864E+02 3.481E+02 1.3522+02 5.1272+401 Te-127m 5.0052-01 3.379E+01 6.709E+01 4.882E+01 1.960E+01 7.680E+00 Te-129 1.020E+01 5.990E+02 9.477E+02 4.228E+02 9.554E+01 2.597E+01 Te-129m 1.603E+00 1.082E+02 2.148E+02 1.5602+02 6.230E+01 2.4252401 CC-AA-309-1001, Rev 2

I I CALCULATION NO. DREO 5-0048 CACLTO NO RO-08IRVSONN.

j REVISION NO. 0 _I _tAGE NO. 46 of 64

!w O4 f6 I Table S (Cont'd)

Post-LOCA Reactor Building Isotopic Inventory - Containment Leakage Post-LOCA Reactor Building Isotopic Inventory (Ci)

Isotope Containment Leakage 0.667 hr 2.0hr 4.0 hrs 8.0 hrs 16brs l 24hrs Te-131m 4.800E+00 3.143E+02 5.956E+02 3.950E+02 1.318E+02 4.2911E+OI

-iTe-132 4.717E+O I3.147E+03 6.138E+03 43 1 OE+03 1.611E+03 5.879E+02

.-131 1.446E+03 I.473E+04 2.593E+04 1.850E+04 7.395E+03 3.001E+03 1-132 1.845E+03 1.591 E+04 1.919E+04 7.301E+03 1.939E+03 7.04 1E+02 1-133 2.877E+03 2.816E+04 4.671 E+04 2.958E+04 9.316E+03 2.979E+03 1-134 1.905E+03 6.792E+03 2.478E+03 7.584E+01 5.584E-02 4.174E-05 1-135 2.572E+03 2.288E+04 3.289E+04 1.564E+04 2.780E+03 5.017E+02 Xe-133 4.203E+03 1.086E+05 3.809E+05 7.496E+05 1.077E+06 1.158E+06 Xe-13S 1.773E+03 4.530E+04 1.448E+05 2.178E3+05 1.768E+05 1.072E+O5 Cs-134 4.262E+02 3.859E+03 6.749E+03 4.847E+03 1.941 E+03 7.600E+02 Cs-136 1.264E+02 1.141 E+03 1.988E+03 1.415E+03 5.569E+02 2.143E+02 Cs-137 2.6231E+02 2.375E+03 4.15S4E+03 2.984E+03 1.195E+03 4.680E+02 Ba-139 1.723E+01 5.95 1E+02 4.321E+02 4.205E+01 3.020E-01 2.117E-03 Ba-140 2.3211E+01 1.562E+03 3.087E+03 2.225E1+03 8.769E+02 3.373E+02 La-140 2.947E-01 4.252E+01 1.7-78E+02 2.690E+02 2.083E+02 1.153E+02 La-141 1.949E-01 1.040E+01 1.451E+O1 5.213E+00 5.103E-01 4.875E-02 La-142 1.566E-01 5.805E+00 4.688E+00 5.646E-01 6.211 E-03 6.669E-05 Ce-141 5.503E-01 3.713E+01 7.3611E+OI 5.338E+01 2.128E+01 8.278E+OO Ce-143 4.984E-01 3.272E+01 6.228E+01 4.166E+01 1.413E+01 4.680E+OO Ce-144 4.538E-01 3.063E+01 6.080E+01 4.421E+01 1.773E+01 6.939E+OO Pr-143 I.955E-0 1.323E+O1 2.641 E+O I 1.942E+01 7.922E+00 3.137E+O0 Nd- 147 8.859E-02 5.960E+00 1.177E+01 8.471 E+00 3.329E+00 1.277E+00 Np-239 6.830E+00 4.537E+02 8.787E+02 6.086E+02 2.214E+02 7.863E+OI Pu-238 2.179E-03 1.471E-0I 2.921 E-0I 2.125E-01 8.527E-02 3.340E-02 Pu-239 1.817E-04 1.227E-02 2.436E-02 1.773E-02 7.121 E-03 2.792E-03 Pu-240 2.466E-04 1.665E-02 3.305E-02 2.404E-02 9.650E-03 3.780E-03 Pu-241 8.258E-02 5.575E+00 1.107E+O1 8.050E+0o 3.231 E+OO 1.266E+00 Am-241 4.860E-05 3.282E-03 6.520E-03 4.748E-03 1.910E-03 7.502E-04 Cm-242 1.126E-02 7.603E-01 1.509E+00 1.097E+00 4.395E-01 1.719E-01 Cm-244 7.992E-04 5.396E-02 1.071 E-0I 7.791 E-02 3.127E-02 1.225E-02 Post-LOCA Reactor Building Isotopic Inventory From RADTRAD Run DRE400CL3 .oO l CC-AA-309- 001, Rev2

ICALCULATION NO. DRE05-0048 IREVISION NO. 0 PAGE NO. 47 of 64 Table 6 Post-LOCA Reactor Building Isotopic Inventory - ESF Leakage Post-LOCA Reactor Building Isotopic Inventory (Ci)

Isotope ESF Lcakage

. 0.667 hr 2.0 hr 4.0 hrs 8.0 hrs 16 hrs 24 hrs 1-131 2.679E+01 3.031 E+02 8.667E+02 1.644E+03 2.378E+03 2.606E+03 1-132 3.403E+01 3.043E+02 5.135E+02 3.01OE+02 4.043E+101 4.098E+00 1-133 5.331E+01 5.795E+02 1.5611E+03 2.630E+03 2.998E+03 2.5911E+03 1-134 3.530E+01 1.398E+02 8.2811E+01 6.744E+00 1.797E-02 3.630E-05 1-135 4.765E+01 4.708E+02 1.100E+03 1.391 E+03 8.947E+02 4.362E+02 Xe-133 1.244E-0 3.941 E+00 2.640E+01 1.082E+02 2.972E+02 4.4511E+02 Xe-135 1.350E+00 3.899E+01 2.266E+02 6.988E+02 1.093E+03 9.3411 E+02 Post-LOCA Reactor Building Isotopic Inventory From RADTRAD Run DRE400CL3 .oO I CC-AA-309-1001, Rev 2

] CALCULATION NO. DREO5-0048 - REVISION NO. 0 ---4 PAGE-NO.-48of 64 Table 7 Post-LOCA Reactor Building Isotopic Inventory - Containment + ESF Leaka es Post-LOCA Reactor Building Isotopic Inventory (Ci) Total Isotope Containment + ESF Leaks e Activity 0.667 hr 2.0 hr 4.0 hrs 8.0 hrs 16 hrs 24 hrs (Cl)

Co-58 9.420E-03 6.357E-O1 1.2611E+00 9.156E-01 3.663 E-0 I 1.430E-01 3.3311E+00

-Co-60 -1.128E-02 -7.614E 1.511E+00 -1.099E+00 4A.12E 1.728E 3.997E+00 Kr-85 3.474E+01 9.007E+b2 3.188E+03 6.41OE+03 9.628E+03 1.082E+04 3.098E+04 Kr-85m 4.862E+02 1.026E+04 2.664E+04 2.885E+04 1.257E+04 4.095E+03 8.289E+04 Kr-87 7.145E+02 8.957E+03 1.066E+04 2.422E+03 4.645E+01 6.665E-01 2.280E+04 Kr-88 1.228E+03 2.299E+04 4.995E+04 3.783E+04 8.064E+03 1.286E+03 1.213E+05 Rb-86 3.713E+O0 3.409E+01 5.944E+01 4.243E+01 1.679E+01 6.494E+00 1.630E+02 Sr-89 1.197E+01 8.072E+02 1.6011E+03 1.162E+03 4.641 E+02 1.81 OE+02 4.227E+03 Sr-90 1.739E+00 1.174E+02 2.331 E+02 1.696E+02 6.805E+01 2.666E+01 6.165E+02 Sr-91 1.447E+01 8.863E+02 1.520E+03 8.261E+02 1.849E+02 4.04-1 E+OI 3.473E+03 Sr-92 1.398E+01 6.71 OE+02 7.986E+02 2.088E+02 1.083E+01 5.482E-01 1.704E+03 Y-90 2.002E-02 2.426E+00 9.302E+00 1.365E+01 1.068E+01 6.062E+00 4.214E+01 Y-91 1.559E-01 .1.068E+01 2.179E+01 1.661 E+O I 7.056E+00 2.844E+00 5.914E+01 Y-92 4.794E-01 1.416E+02 4.948E+02 3.502E+02 5.202E+01 5.473E+00 1.045E+03 Y-93 1.857E-01 1.144E+01 1.979E+01 1.094E+01 2.536E+00 5.737E-01 4.547E+01 Zr-95 2.190E-01 1.478E+01 2.930E+01 2.128E+01 8.509E+00 3.321E+00 7.7411E+01 Zr-97 2.157E-01 1.379E+01 2.522E+01 1.557E+01 4.SOOE+00 1.270E+00 6.056E+01 Nb-95 2.201 E-0 I 1.486E+01 2.949E+01 2.145E+01 8.609E+00 3.372E+00 7.801E+01 Mo-99 3.134E+00 2.086E+02 4.056E+02 2.829E+02 1.044E+02 3.759E+01 1.042E+03 Tc-99m 2.7642+00 1.864E+02 3.6842+02 2.633E+02 1.012E+02 3.764E+01 9.597E+02 Ru-103 2.656E+00 1.7912E+02 3.5502+02 2.575E+02 1.027E+02 4.001E+01 9.370E+02 Ru-105 1.685E+00 9.238E+01 1.342E+02 5.228E+01 6.018E+00 6.761E-01 2.872E+02 Ru-106 1.132E+00 7.642E+01 1.517E+02 1.103E+02 4.424E+01 1.732E+01 4.0111E+02 Rh- IOS 1.776E+00 1.195E+02 2.341 E+02 1.629E+02 5.761E+01 1.948E+01 5.953E+02 Sb-127 3.678E+00 2.4582+02 4.807E+02 3.394E+02 1.283E+02 4.732E201 1.245E+03 Sb-129 9.831 E+00 5.359E+02 7.718E+02 2.955E+02 3.286E+01 3.566E+00 1.649E+03 Tc-127 3.672E+00 2.468E+02 4.8642+02 3.4812E+02 1.3522+02 5.1272+01 1.271 E+03 Te-127m 5.0052-01 3.379E+01 6.709E+01 4.882E+01 1.960E+01 7.6802+00 1.775E+02 Te-129 1.0202+01 5.990E+02 9.477E+02 4.228E+02 9.554E+01 2.597E+01 2.1012E+03 Te-129m 1.6032+00 1.082E+02 2.148E+02 1.560E+02 6.230E+01 2.425E+01 5.672E+02 CC-AA-309-1001, Rev 2

I CALCULATION NO. DRE05-0048 I REVISION NO. 0 l PAGE NO. 49 of 64 Table 7 (Cont'd)

Post-LOCA Reactor Building Isotopic Inventory - Containment + ESF Leakages Post-LOCA Reactor Building Isotopic Inventory (Ci) Total Isotope .Containment + ESF Leaka e Activity 0.667 hr 2.0 hr 4.0 hrs 8.0 hrs 16 hrs 24 hrs (C)

Te-13im 4.800E+o0 3.143E+02 5.956E+02 3.950E+02 1.318E+02 4.291 E+O I 1.484E+03 Te-132 4.717E+01 3.147E+03 6.138E+03 4.31 0E+03 1.611 E+03 5.879E+02 1.584E+04 1-131 1.473E+03 1.503E+04 2.680E+04 2.014E+04 9.772E+03 5.607E+03 7.883E+04 1-132 1.879E+03 1.621 E+04 1.971 E+04 7.602E+03 1.980E+03 7.082E+02 4.809E+04 1-133 2.931 E+03 2.874E+04 4.827E+04 3.220E+04 1.231 E+04 5.570E+03 1.300E+05 1-134 1.940E+03 6.932E+03 2.560E+03 8.258E+01 7.381 E-02 7.804E-05 1.15S2E+04 1-135 2.619E+03 2.335E+04 3.399E+04 1.703E+04 3.675E+03 9.379E+02 8.1611E+04 Xe-133 4.203E+03 1.086E+05 3.809E+05 7.497E+05 1.077E+06 1.158E+06 3.479E+06 Xe-135 1.775E+03 4.534E+04 1.450E+05 2.185E+05 1.779E+05 1.082E+05 6.967E+05 Cs-134 4.262E+02 3.859E+03 6.749E+03 4.847E+03 1.941 E+03 7.600E+02 1.858E+04 Cs-136 1.264E+02 1.141E+03 1.988E+03 1.415E+03 5.569E+02 2.143E+02 5.442E+03 Cs-137 2.623E+02 2.375E+03 4.154E+03 2.984E+03 1.195E+03 4.680E+02 1.144E+04 Ba-139 1.723E+01 5.951 E+02 4.3211E+02 4.205E+01 3.020E-01 2.117E-03 1.087E+03 Ba-140 2.321E+01 1.562E+03 3.087E+03 2.225E+03 8.769E+02 3.373E+02 8.11 IE+03 La-140 2.947E-01 4.252E+01 1.778E+02 2.690E+02 2.083E+02 1.153E+02 8.133E+02 La-141 1.949E-01 1.040E+01 1.451E+0I 5.213E+O0 5.103E-01 4.875E-02 3.088E+01 La-142 1.566E-01 5.805E+00 4.688E+00 5.646E-01 6.211 E-03 6.669E-05 1.122E+O01 Ce-141 5.503E-01 3.713E+01 7.3611E+01 5.338E+01 2.128E+01 8.278E+00 1.942E+02 Ce-143 4.984E-01 3.272E+01 6.228E+01 4.166E+01 1.413E+01 4.680E+00 1.560E+02 Ce-144 4.538E-01 3.063E+01 6.080E+01 4.421 E+OI 1.773E+01 6.939E+00 1.608E+02 Pr-143 1.95SE-01 1.323E+01 2.641 E+O I 1.942E+01 7.922E+00 3.137E+00 7.032E+01 Nd-147 8.859E-02 5.960E+00 1.177E+01 8A71 E+0O 3.329E+00 1.277E+00 3.089E+01 Np-239 6.830E+00 4.537E+02 8.787E+02 6.086E+02 2.214E+02 7.863E+01 2.248E+03 Pu-238 2.179E-03 1.471E-0I 2.9211E-O1 2.125E-01 8.527E-02 3.340E-02 7.725E-01 Pu-239 1.817E-04 1.227E-02 2.436E-02 I.773E-02 7.121 E-03 2.792E-03 6.445E-02 Pu-240 2.466E-04 1.665E-02 3.305E-02 2.404E-02 9.650E-03 3.780E-03 8.742E-02 Pu-241 8.258E-02 5.575E+00 1.107E+01 8.050E+00 3.231 E+00 1.266E+00 2.927E+O01 Am-241 4.860E-05 3.282E-03 6.520E-03 4.748E-03 1.910E-03 7.502E-04 1.726E-02 Cm-242 1.126E-02 7.6031E-01 1.509E+00 1.097E+00 4.395E-01 1.719E-01 3.988E+O0 Cm-244 7.992E-04 5.396E-02 1.071E-01 7.791 E-02 3.127E-02 1.225E-02 2.833E-01 Containment Leakage RB Inventory From Table 5 ESF Leakage RB Inventory From Table 6 I CC-AA-309-1 001, Rev 2

-CALCULATION NO. DREOS-0048 CACLTO NO DRO-04

- 2 - -ISION NO. 0 REIINN.

] _VAGENO.'50 of 64 AEN;5 f6 I Table 8 Comparison of Post-LOCA 0-24 hrs Isotopic Activity In Reactor Building 0-24 hrs Isotonic Activitv In Reactor Buildinv (Cl Isotope Quad Cities Dresden Isotope Quad Cities Dresden Units'l & 2 Units 2 & 3 Units I & 2 Units 2 & 3

_ _A _ B A ___B Co-58 3.398E+00 3.331E+0O Te-131m 1.51 OE+03 i.484E+03 Co-60 4.078E+00 3.997E+OO Te-132 1.614E+04 1.584E+04 Kr-85 3.155E+04 3.098E+04 1-131 8.041 E+04 7.883E+04 Kr-85m 8.360E+04 8.289E+04 1-132 4.8611E+04 4.809E+04 KI-87 2.286E+04 2.280E+04 1-133 1.322E+05 1.300E+05 Kr-88 1.220E+05 1.213E+05 1-134 1.156E+04 1.152E+04 Rb-86 i.663E+02 1.630E+02 1-135 8.254E+04 8.161 E+04 Sr-89 4.312E+03 4.227E+03 Xe-133 3.542E+06 3.479E+06 Sr-90 6.290E+02 6.165E+02 Xe-135 7.061 E+05 6.967E+05 Sr-91 3.517E+03 3.473E+03 Cs-134 1.896E+04 i.858E+04 Sr-92 1,715E+03 1.704E+03 Cs-136 5.550E+03 5.442E+03 Y-90 4.383E+01 4.214E+0I Cs-137 1.167E+04 1.144E+04 Y-91 6.041 E+O I 5.914E+01 Ba-139 1.09 1E+03 1.087E+03 Y-92 1.060E+03 1.045E+03 Ba-140 8.272E+03 8.11 E+03 Y-93 4.607E+01 4.547E+01 La-140 8.460E+02 8.133E+02 Zr-95 7.897E+01 7.741E+01 La-141 3.114E+01 3.088E+01 Zr-97 6.150E+01 6.056E+01 La-142 1.127E+01 1.122E+0I Nb-95 7.958E+10I 7.801E+01 Ce-141 1.981E+02 1.942E+02 Mo-99 1.062E+03 1.042E+03 Ce-143 1.587E+02 1.560E+02 Tc-99m 9.784E+02 9.597E+02 Ce-144 1.640E+02 1.608E+02 Ru-103 9.558E+02 9.370E+02 Pr-143 7.176E+01 7.032E+O1 Ru-105 2.898E+02 2.872E+02 Nd-147 3.151 E+O I 3.089E+01 Ru-106 4.092E+02 4.011 E+02 Np-239 2.290E+03 2.248E+03 Rh-105 6.062E+02 5.953E+02 Pu-238 7.8811E-01 7.725E-O01 Sb-127 1.269E+03 1.245E+03 Pu-239 6.576E-02 6.445E-02 Sb-129 1.664E+03 1.649E+03 Pu-240 8.919E-02 8.742E-02 Te-127 1.296E+03 I.271E+03 Pu-241 2.986E+01 2.927E+0I Te-127m 1.81 IE+02 1.775E+02 Am-241 1.761 E-02 1.726E-02 Te-129 2.126E+03 2.101 E+03 Cm-242 4.069E+00 3.988E+00 Te-129m 5.787E+02 5.672E+02 Cm-244 2.890E-01 2.833E-O01 A From Reference 9.20, Table 6 & B From Table 7 I CC-AA-309-1001, Rev 2 l

..,--I-CALCULATION.NODREOS-0048 CACLTONODE504 4 I-REVISIONNO.0 RVSO O0jAEN.5

- - --1 -PAGE NO. 51 of 64 f6 JIl Table 9 Post-LOCA Elemental Iodine Inventory Transported to the Environment Due to Post-LOCA MSIV Leakage Time - Failed MS Line Intact MS Line 1 Intact MIS Line 2 Total Time MSIV Cumulative Cumuiative Cumulative Cumulative Interval Elem. Iodine Elem. Iodine Elem. Iodine Elem. Iodine Elem. Iodine Transported Transported Transported Transported Transported to Environment to Environment to Environment to Environment to Environment (hrs) (atoms) (atoms) (atoms) (atoms) (hrs) (atoms)

IAI FBI ICI IA+B+C1 0.6667 1.1422E+16 1.3810E+15 1.7915E+14 1.2982E+16 2 1.9714E+17 5A977E+16 8.6096E+15 2.6073E+17 0.6667 to 2 2.4774E+17 3.615 4.0029E+17 1.4958E+17 2.6703E+16 5.7657E+17 8 7.3746E+17 3.6490E+17 9.2433E+16 1.1948E+18 2 to 8 9.3407E+17 24 1.0209E+18 5.1323E+17 2.1569E+17 1.7498E+18 8 to 24 5.5503E+17 96 I.1I195E+18 5.5808E+17 2.6193E+17 1.93952+18 24 to 96 1.8969E+17 720 1.2995E+18 6.4809E+17 *3.0710E+17 2.2547E+18 96 to 720 3.1518E+17 A, B & C From RADTRAD Run DRE400MS3I.oO output file Table 10 Post-LOCA Total Elemental Iodine Inventory On CR Charcoal Filter @ 720 firs Due to Post-LOCA MSIV Leakage Time MSIV X/Q Time Volume HVAC Charcoal Filter Interval Elem. Iodine iNSIV to Conversion Conversion Inflow Filter Inventory Transported CR rate Efficiency Elem. Iodine to Environment (hrs) (atoms) (sec/m3) (min/sec) (m3/ft3) (tt3/min) (fraction) (atoms)

_[_Al IB1 ICI [DI IEJ IF] [A*B*C*D*E*F1 0.6667 to 2 2.4774E+17 1.30E-03 0.01667 0.02832 1800 0.99 2.709E+14 2 to 8 9.3407E+17 1.06E-03 0.01667 0.02832 1800 0.99 8.328E+14 8 to 24 5.5503E+17 4.49E-04 0.01667 0.02832 1800 0.99 2.096E+14 24 to 96 1.8969E+17* 2.96E-04 0.01667 0.02832 1800 0.99 4.723E+13 96 to 720 3.1518E+17 2.44E-04 0.01667 0.02832 1800 0.99 6.468E+13 Total- 1A25E+15 l CC-AA-309-1001. Rev 2

ICALCULATIO:N NO. DRE:05-0048 I R lREVISION NO. 0 JPAGE NO. 52 of 64 Table 11 Post-LOCA Organic Iodide Inventory Transported to the Environment Due to Post-LOCA MSIV Leakage Time Failed MS Line Intact MS Line I Intact NIS Line 2 Total Time MSIV Cumulative Cumulative Cumulative Cumulative Interval Org. Iodide Org. Iodide Org. Iodide Org. Iodide Org. Iodide Transported Transported Transported Transported Transported to Environment to Environment to Environment to Environment to Environment (hrs) (atoms) (atoms) (atoms) (atoms) (hrs) (atoms)

[Al 13 BI CI iA+B+CI 0.6667 1.1320E+15 2.5436E+14 3.2803E+13 1.4192E+15 2 3.5093E+16 1.7432E+16 2.6480E+15 5.5173E+16 0.6667 to 2 5.3754E+16 3.615 1.0430E+17 6.6345E+16 I.IISSE+16 1.8180E+17 8 4.2853E+17 3.7018E+17 8.0274E+16 8.7898E+17 2 to 8 8.2381E+17 24 2.0672E+1 8 2.0467E+1 8 7.0037E+17 4.8143E+18 8 to 24 3.9353E+1 8 96 8.0892E+18 8.0759E+18 3.6664E+18 1.9832E+19 24 to 96 1.5017E+19 720 2.1236E+19 2.1224E+19 1.0262E+19 5.2722E+19 96 to 720 3.2891E+19 A,B & C From RADTRAD Run DRE400NIS31.oO output file Table 12 Post-LOCA Total Organic Iodide Inventory On CR Charcoal Filter @ 720 Hrs Due to Post-LOCA AISIV Leakage Time MSIV X/Q Time Volume HVAC Charcoal Filter Interval Organic Iodide MSIV to Conversion Conversion inflow Filter Inventory Transported CR rate Efficiency Organic Iodide to Environment (hrs) (atoms) (sec/m3) (min/sec) (m3/ft3) (ft3/min) (fraction) (atoms)

[Al IIL ICl. _ ID] [El J17i [A*B*C*D*E*FI 0.6667 to 2 5.3754E+16 1.30E-03 0.01667 0.02832 1800 0.99 5.878E+13 2 to 8 8.2381E+17 1.06E-03 0.01667 0.02832 1800 0.99 7.345E+14 8 to 24 3.9353E+18 4.49E-04 0.01667 0.02832 1800 0.99 1.486E+15 24 to 96 1.5017E+19 2.96E-04 0.01667 0.02832 1800 0.99 3.739E+I5 96 to 720 3.2891 E+19 2.44E-04 0.01667 0.02832 1800 0.99 6.750E+I5 Total - 1.277E+16 CC-AA-309-1001 Rev2

-- I CALCULATION N-0,I1REQ"0048 _ REYISION.NO. 0 -_JPAGENO. 53.of 64 I Table 13 Post-LOCA Aerosol Inventory Transported to the Environment Due to Post-LOCA MSIV Leakage Time Failed MS Line Intact hMS Line I Intact MS Line 2 Total Time MSIV Cumulative Cumulative Cumulative Cumulative Interval Aerosols Aerosols Aerosols Aerosols Aerosols Transported Transported Transported Transported Transported to Environment to Environment to Environment to Environment to Environment (hrs) (kg) (kg) (kg) (kg) (hr5) (kg)

AL1 B1 [CL [A+B+CI 0.6667 4.3719E-06 1.4919E-07 4.6206E-09 4.5257E-06 2 1.1614E-04 8.8456E-06 3.2488E-07 1.2531 E-04 0.6667 to 2 1.2078E-04 3.615 3.0129E-04 3.0630E-05 1.2602E-06 3.3318E-04 8 7.0887E-04 _ 1.0261E-04 5.9502E-06 8.1743E-04 2 to 8 6.9212E-04 24 1.0788E-03 1.6261E-04 1.6832E-05 1.2582E-03 8 to 24 4.4081 E-04 96 1.10SI E-03 1.6447E-04 L.9641 E-05 1.2892E-03 24 to 96 3.0969E-05 720 1.1051 E-03 1.6447E-04 1.9643E-05 1.2892E-03 96 to 720 2.0000E-09 A, B & C From RADTRAD Run DRE400MS3 .oO output file Table 14 Post-LOCA Total Aerosol Inventory On CR HEPA Filter @ 720 Hlrs Due to Post-LOCA NISIV Leakage Time MSIV X/Q Time Volume HVAC HEPA Filter Interval Aerosols MSIV to Conversion Conversion inflow Filter Inventory Transported CR rate Efficiency Aerosols to Environment (hrs) (kg) (sec/m3) (minIsec) (m3/ft3) (ft3/1min) (fraction) (kg)

[Al I13l ICI IDI IEJ 1171 [A*B*C*D*E*F 0.6667 to 2 1.2078E-04 1.30E-03 0.01667 0.02832 1800 0.99 1.321E-07 2 to 8 6.9212E-04 1.06E-03 0.01667 0.02832 1800 0.99 6.171E-07 8 to 24 4.4081E-04 4.49E-04 0.01667 0.02832 1800 0.99 1.66SE-07 24 to 96 3.0969E-05 2.96E-04 0.01667 0.02832 1800 0.99 7.710OE-09 96 to 720 2.0000E-09 2.44E-04 0.01667 0.02832 1800 0.99 4.105E-13 Total - 9.233E-07 l CC-AA-309-1001, Rev2

I I CALCULATION NO. DRE05-0048 CALCULATION NO. DREOS-0048 I1 REVISION NO. 0 REVISION NO.0 ]

_I PAGE NO.'54 of 64 PACE NO.54 of 64 I Table 15 Conversion of Iodine Activity Into Iodine Atom RB Region A 0.5 hr lodine Isotopic Isotope Activity Atoms Atoms Per Iodine (Curie) (Curie) Fraction

-_-A 1 --- B -_ - C - B___-_D_ - B/1 1-131 8.0522E+02 2.9858E+19 3.708E+16 7.693E-O1 1-132 L0541E+03 4.6588E+17 4.420E+14 1.200E-02 1-133 1.6103E+03 6.4364E+18 3.997E+15 1.658E-01 1-134 1.2097E+03 2.0380E+17 1.685E+14 5.251E-03 1-135 1.4565E+03 1.8501 E+18 1.270E+15 4.767E-02 Total 3.8811E+l19 1.000E+O0 Ai & B, From RADTRAD Run DRE400CL3 .oO output file _

a 0.5 hr from Reactor Building Compartment Nuclide Inventory Table 16 Post-LOCA MSIV Leakage Iodine Activity Deposited on CR Charcoal Filter Iodine Fraction Elemental & Iodine Iodine Isotope Atoms Per Of Iodine Organic lodin Atoms on Activity Curie Atoms On CR Charcoal CR Charcoal CR Charcoal Filter Filter 720 Hrs At 720 irs At 720 Hrs Cl A B C Di -B,

• C E - D /A 1-131 3.708E+16 7.693E-01 .1.4195E+16 1.092E+16 2.945E-01 1-132 4.420E+14 1.200E-02 1.704E+14 3.855E-01 1-133 3.997E+15 1.658E-01 2.354E+15 5.889E-01 1-134 1.685E+14 5.251E-03 7.453E+13 4.424E^-01 1-135 1.270E+15 4.767E-02 6.766E+14 5.327E-0O Total Iodine Sump Atoms/Activity 1.420E+16 2.244E+00 A4 & Bi From Table 14 C From Section 7.11 (Table 10 + Table 12 atom inventories)

I CC-AA-309-1001, Rev 2

I CALCULATION NO. DREO5-0048 I REVISION NO. 0 l PAGE NO. 55 of 64 Table 17 Rclatlonship of Aerosol Mass and Activity CR Region a. 0.6667 hr Aerosol Isotopic Isotope Activity Mass Mass Per Cl Aerosol (Curie) (kg) (kg/Ci) Fraction Ai B.i -C - B. /At Di - 1391 Co-58 9.420E-03 2.963E-10 3.145E-08 8.802E-08 Co-60 1.128E-02 9.977E-09 8.846E-07 2.964E-06 Rb-86 3.773E+00 4.636E-08 1.229E-08 1.378E-05 Sr-89 1.197Ef01 4.119E-07 3.442E-08 1.224E-04 Sr-90 1.739E+00 1.275E-05 7.331E-06 3.789E-03 Sr-91 1.447E+01 3.991E-09 2.759E-10 1.186E-06 Sr-92 1.398E+01 1.1 12E-09 7.956E-11 3.304E-07 Y-90 2.002E-02 3.680E-1 I 1.838E-09 1.093E-08 Y-91 1.559E-01 6.356E-09 4.07SE-08 1.888E-06 Y-92 4.794E-01 4.982E-1 I 1.039E-10 1.480E-08 Y-93 1.857E-O1 5.565E-1 I 2.997E-10 1.653E-08 Zr-95 2.190E-01 1.019E-08 4.655E-08 3.028E-06 Zr-97 2.157E-01 1.129E-10 5.231 E-10 3.353E-08 Nb-95 2.2011E-0I 5.628E-09 2.557E-08 1.672E-06 Mo-99 3.134E1+00 6.534E-09 2.085E-09 1.9411E-06 Tc-99m 2.764E+OO 5.256E-10 1.902E-10 1.562E-07 Ru-103 2.656E+00 8.228E-08 3.098E-08 2.445E-05 Ru-105 1.685E+00 2.507E1-0 1.488E-10 7.447E-08 Ru-106 1.132E+00 3.384E-07 2.989E-07 I.005E-04 Rh-lOS 1.776E+00 2.104E-09 1.185E-09 6.252E-07 Sb-127 3.678E+00 1.377E-08 3.745E-09 4.092E.06 Sb-129 9.831 E+OO 1.748E-09 1.778E-10 5.194E-07 Tc-127 3.672E+00 _.392E-09 3.789E- 10 4.134E-07 Te-127m 5.005E-0I 5.306E-08 1.060E-07 1.576E-OS Te-129 1.020E+01 4.869E-10 4.775E-1I 1.447E-07 Te-129m 1.603E+00 5.320E-08 3.319E-08 I.5811E-05 CC-AA-309-1001, Rev 2

_l CALCULATION NO. DREOS-0048 -- --- I _gliVSION NO. 0 -- j YAGE NO. 56 of 64 -I Table 17 (Cont'd)

Relationship of Aerosol Mass and Activity CR Region ) 0.6667 hr Aerosol Isotopic Isotope Activity Mass Mass Per Ci Aerosol (Curie) (kg) (kg/CI) Fraction Al B C,- Bf /Al DI - B/ZB Te-131m 4.800E+00 6.020E-09 1.254E-09 1.789E-06 Te-132 4.717E+01 1.554E-07 3.294E-09 4.617E-05 Cs-134 4.262E+02 3.294E-04 7.729E-07 9.788E-02 Cs-136 1.264E+02 1.725E-06 1.364E-08 5.125E-04 Cs-137 2.623E+02 3.015E-03 1.150IE-05 8.958E-OI Ba-139 1.723E+01 1.054E-09 6.1131E-11 3.130E-07 Ba-140 2.3211E+O1 3.170E-07 1.366E-08 9.418E-05 La-140 2.947E-01 5.302E-10 1.799E-09 1.575E-07 La-141 1.949E-01 3.447E- 1I 1.768E-10 1.024E-08 La-142 1.566E-01 1.094E-1I 6.986E-11 3.250E-09 Ce-141 5.503E-01 1.931E-08 3.510E-08 5.738E-06 Cc- 143 4.984E-01 7.506E-10 1.506E-09 2.230E-07 Ce-144 4.538E-01 1.423E-07 3.135E-07 4.227E-05 Pr- 143 1.955E-01 2.903E-09 1.485E-08 8.624E-07 Nd-147 8.859E-02 I.0953E-09 L.236E-08 3.254E-07 Np-239 6.830E+00 2.944E-08 4.311 E-09 8.748E-06 Pu-238 2.179E-03 1.273E-07 5.841 E-05 3.782E.05 Pu-239 1.8177E-04 2.923E-06 1.609E-02 8.685E-04 Pu-240 2.466E-04 1.082E-06 4.389E-03 3.216E-04 Pu-241 8.258E-02 8.017E-07 9.708E-06 2.382E-04 Am-241 4.860E-05 1.416E-08 2.914E-04 4.207E-06 Cm-242 1.126E-02 3.399E-09 3.017E-07 1.01 OE-06 Cm-244 7.992E-04 9.878E-09 1.236E-05 2.935E-06 Total 3.366E-03 1.OOOE+OO A; & Bi From RADTRAD Run DRE400CL3 1.oO output file @

0.6667 hr from Reactor Building Compartment Nuclide Inventory I CC-AA-309-1001, Rev2 I

CALCULATION NO. DRE0"048 I REVISION NO. 0 PAGE NO. 57 or 64I Table 18 Post-LOCA Total Aerosol Isotopic Activity On CR HEPA Filter @ 720 Hrs Post-LOCA NISIV Leakage Aerosol Fraction Total Aeros I Isotopic Isotope Mass Per Cl of CR Filter Aerosol Mass Aerosol Activity Aerosol Aerosol Mass On CR Filter On CR Filter At 720 Hr At 720 Hr At 720 llr (kg/Ci) (kg) (kg) (Ci)

AL B, C D B *C E 1-DI/A 1 Co-58 3.145E-08 8.802E-08 9.233E-07 2.470E-17 7.853E-10 Co-60 8.846E-07 2.964E-06 8.317E-16 9.402E-10 Rb-86 1.229E-08 1.378E-05 3.865E-15 3.145E-07 Sr-89 3.442E-08 1.224E-04 3.434E-14 9.975E3-07 Sr-90 7.331E-06 3.789E-03 1.063E-12 1.450E-07 Sr-91 2.759E-10 1.186E-06 3.327E-16 1.206E-06 Sr-92 7.956-1I 3.304E-07 9.270E-17 1.165E-06 Y-90 1.838E-09 1.093E-08 3.068E-18 1.669E-09 Y-91 4.078E-08 1.888E-06 5.298E-16 1.299E-08 Y-92 1.039E-10 1.480E-08 4.153E-18 3.996E-08 Y-93 2.9972-10 1.653E-08 4.639E-18 1.548E-08 Zr-95 4.6552-08 3.028E-06 8.497E-16 1.826E-08 Zr-97 5.2312-10 3.353E-08 9.408E-18 1.798E-08 Nb-95 2.557E-08 1.672E-06 4.692E-16 1.835E-08 Mo-99 2.085E-09 1.941E-06 5.447E-16 2.613E-07 Tc-99m 1.9022-10 1.5624-07 . 4.382E-17 2.304E-07 Ru-103 3.098E-08 2.445E-05 6.859E-15 2.214E-07 Ru-105 1.488E-10 7.447E-08 2.090E-17 1.405E-07 Ru-106 2.9892-07 1.005E-04 2.821E-14 9.437E-08 Rh-105 1.185E-09 6.2522-07 1.754E-16 1.481E-07 Sb-1 27 3.745E-09 4.092E-06 1.148E-15 3.066E-07 Sb-129 1.778E-10 5.194E-07 IA57E-16 8.196E-07 Te-127 3.789E-10 4.134E-07 1.160E-16 3.061E-07 Te-127m 1.060E-07 1.576E-05 4.423E-15 4.172E-08 Te-129 4.7752-11 1.447E-07 4.059E-17 8.5002-07 Te-129m 3.3192-08 1.581E-05, 4.435E-15 1.336E-07 CC-AA-309-1001,Rev2

_ CALCULATION NO. DRE05-0048 --.-- I.REVISIONNO.0 - JPAGE NO. 58 of 64 I Table 18 Post-LOCA Total Aerosol Isotopic Activity On CR HEPA Filter @ 720 lIrs Post-LOCA NISIV Leakage Aerosol Fraction Total Aerosol Isotopic Isotope Alass Per Cil of CR Filter Aerosol Mass Aerosol Activity Aerosol Aerosol Mass On CR Filter On CR Filter At 720 Hr At 720 Hir At 720 llr (kg/Ci) (kg) (kg) (Ci)

Ai .C D, -BI

• C Ef-DI / Al Te-131m 1.254E-09 1.789E-06 9.223E-07 5.018E-16 4.002E-07 Te-132 3.294E-09 4.6177E-05 I .295E-14 3.933E-06 Cs-134 7.729E-07 9.788E-02 2.746E-1 I 3.5531E-05 Cs-136 1.364E-08 5.125E-04 1.438E-13 1.054E-05 Cs-137 1.150E-05 8.958E-01 2.514E-10 2.186E-05 Ba-139 6.113E- II 3.130E-07 8.783E1-17 1.437E-06 Ba-140 1.366E-08 9.418E-05 2.642E-14 1.935E-06 La-140 1.799E-09 l.575E-07 4.420E-17 2.457E-08 La-141 1.768E-10 1.024E-08 2.873E-18 1.625E-08 La-142 6.986E-11 3.250E-09 9.118E-19 1.305E-08 Ce-141 3.510E-08 5.738E-06 1.610E-15 4.587E-08 Ce-143 1.506E-09 2.230E-07 6.257E-17 4.15S5E-08 Ce-144 3.135E-07 4.227E-05 1.186E-14 3.783E-08 Pr-143 1.485E-08 8.624E-07 2.420E-16 1.630E-08 Nd-147 1.236E-08 3.254E-07 9.129E-17 7.385E-09 Np-239 4.311 E-09 8.748E-06 2.454E-15 5.694E-07 Pu-238 5.841 E-05 3.782E-05 1.061 E-14 1.817E-10 Pu-239 1.609E-02 8.685E-04 2.437E-13 1.515E-I I Pu-240 4.389E-03 3.216E-04 9.023E-14 2.056E-11 Pu-241 9.708E-06 2.382E-04 6.683E-14 6.884E-09 Am-241 2.914E-04 4.207E-06 1.1811E-15 4.052E-12 Cm-242 3.017E-07 1.0 IOE-06 2.833E-16 9.390E-10 Cm-244 1.236E-05 2.935E-06 8.235E-16 6.662E-1 I Ai & Bi From Table 10 C From Section 7.11 (Table 14 kilogram inventory)

I CC-AA-309-1001, Rev 2 I

I

_l-JCA~LCULATION NO. DRE05-0048 CALCULATION NO. DREOS-0048 2

. -J REYISION NO. 0 JIEVISION NO.0 ..PAGENO.59 of 2--=lAGE-NO.59 -64 of64 I 11.0 FIGURES Figure 1: Containment & ESF Leakage RADTRAD Nodalization CC-AA-309-1001, Rev 2

-- _CALCULATJONNO-DRE05-0048 - REVJSION NO. 0 _ SI-AGE-NO. 60 of 64 I Figure 2: MISIV Leakage RADTRAD Nodalization CC-AA-309-1 001, Rev 2

I CALCULATION NO. DRE05-0048 REVISION -NO.0-

-I lI PAGE NO. 61 Or 64-I Figure 3 - Dresden Control Room RADTRAD Nodalization I

i CC-AA-309-100l,Rev2

I CALCULATION NO. DRE05-0048 REVISION NO. 0 PAGE NO. 62 of 64 DP = Dose Point Figure 4: Elevation View of Containment Shine Shielding Geometry Looking @ West I CC-AA-309-1001,Rev2

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1 CALCULATION NO-DREO5-0048 .- . lREVISION NO. 0 -------- PAGE NO.63 of64 I Figure 5: Plan View of Containment Shine Shielding Geometry I CC-AA-309-1001, Rev 2

ICALCULATION NO. DRE05-0648 I REVISION NO. 0 I PAGE NO. 64 of 64 l 12.0 AFFECTED DOCUMENTS Upon approval of the Alternative Source Term Licensing Change Request (LCR), the following documents will be changed:

UFSAR Information To Be Revised (Ref. 9.18):

UFSAR Section 15.6.5 UFSAR Table 15.6.9 UFSAR Table 15.6.1 Oa '

Document To Be Superseded SWEC Calculation No. DREO1-0040, Rev 0, Site Boundary and Control Room Doses following a Loss of Coolant Accident using Alternative Source Terms (Ref. 9.19).

13.0 ATTACHMENTS Diskettes with the various electronic files.

Calculation No: DRE05-0048, Rev 0 (PDF File)

Comment Resolutions I CC-AA-309-1001, Rev 2