ML042740509

Post-LOCA Eab, LPZ, & CR Doses - Alternative Source Term (AST)

ML042740509
Person / Time
Site:	Salem, Hope Creek
Issue date:	02/24/2004
From:	Drucker M, Gita Patel NUCORE
To:	Office of Nuclear Reactor Regulation
References
LR-N04-0413 S-C-ZZ-MDC-1945, Rev. OIR1
Download: ML042740509 (79)

Text

1NC.DE-AP.ZZ-0002(Q)

CALC NO.: S-C-ZZ-MDC-1945 CALCULATION COVER SHEET Page 1 of eW 78 4 REVISION: OIRI CALC. TITLE:

Post-LOCA EAB, LPZ, & CR Doses -Alternative Source Term (AST)

1. SHTS (CALC): Ii

1. 7t ATT I # SHTS: Ill I

1. IDWV50.59 SHTS:

313/3

1. TOTAL SHTS:

1-94 35,

CHECK ONE:

1 El FINAL 0 INTERIM (Proposed Plant Change)

E. FINAL (Future Confirmation Req'd)

El VOID SALEM OR HOPE CREEK:

El Q - LIST 0 IMPORTANT TO SAFETY E1 NON-SAFETY RELATED HOPE CREEK ONLY:

[IQ nQs E]Qsh OF OR E STATION PROCEDURES IMPACTED, IF SO CONTACT RELIABILITY ENGINEER El CDs INCORPORATED (IF ANY):

DESCRIPTION OF CALCULATION REVISION (IF APPL.):

Revision to incorporate the containment spray interruption of 10 minutes during transition from the injection to recirculation phase.

PURPOSE:

The purpose of this calculation is to determine the post-LOCA doses at the Exclusion Area Boundary (EAB), Low Population Zone (LPZ), and Control Room (CR) using the Alternative Source Term (AST), guidance in the Regulatory Guide (RG) 1. 183, RADTRAD3.02 computer code, and Total Effective Dose Equivalent (TEDE) dose criteria for the following post-LOCA release paths:

1.

Containment Leakage

2.

Engineered Safety Feature (ESF) Leakage

3.

Containment Vacuum Relief Line Release

4.

Back-leakage to The Refueling Water Storage Tank (RWST)

The CR dose is evaluated using a CR unfiltered inleakage of 150 cfln, which is greater than the nominal 100 cfm inleakage measured by the baseline Tracer Gas Testing.

CONCLUSIONS:

The post-LOCA results presented in Section 7.0 indicate that the EAB, LPZ, and CR doses due to a LOCA are within their allowable TEDE dose limits.

INulaComnRvso9 I Nuclear Common Revision 9 1

CALCULATION CONTINUATION SHEET SHEET 2 of 78 CALC. NO.: S-C-ZZ-MDC-1945

REFERENCE:

LCR S03-05 G. Patel/NUCORE, ORIGINATOR, DATE REV:

02/24/04 0

M Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/26/04 REVISION HISTORY Revision DESCRIPTION OIRO Initial issue OIRI Revised to incorporate the containment spray interruption of 10 minutes during transition from the injection to recirculation phase.

Nu l a Co mo Re iio I Nuclear Common Revision 9 1

CALCULATION CONTINUATION SHEET SHEET 3 of 78 CALC. NO.: S-C-ZZ-MDC-1945

REFERENCE:

LCR S03-05 G. Patel/NUCORE, ORIGINATOR, DATE REV:

02/24/04 0

M Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/26/04 SHEET REVISION INDEX SHEET REV SHEET REV 1

0 42 0

2 0

43 0

3 0

44 0

4 0

45 0

5 0

46 0

6 0

47 0

7 0

48 0

8 0

49 0

9 0

50 0

10 0

51 0

11 0

52 0

12 0

53 0

13 0

54 0

14 0

55 0

15 0

56 0

16 0

57 0

17 0

58 0

18 0

59 0

19 0

60 0

20 0

61 0

21 0

62 0

22 0

63 0

23 0

64 0

24 0

65 0

25 0

66 0

26 0

67 0

27 0

68 0

28 0

69 0

29 0

70 0

30 0

71 0

31 0

72 0

32 0

73 0

33 0

74 0

34 0

75 0

35 0

76 0

36 0

77 0

37 0

78 0

38 0 3 0

39 0

40 0

41 0

I Nuclear Common Revision 9 l I Nuclear Common Revision 9

I CALCULATION CONTINUATION SHEET ISHEET 4 of 78 CALC. NO.: S-C-ZZ-MDC-1945

REFERENCE:

LCR S03-05 G. Patel/NUCORE, ORIGINATOR, DATE REV:

02/24/04 0

M Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/26/04 TABLE OF CONTENTS Section Cover Sheet Revision History Sheet Revision Index Table of Contents 1.0 Purpose

2.0 Background

3.0 Analytical Approach 4.0 Assumptions 5.0 Design Inputs 6.0 Calculations 7.0 Results 8.0 Conclusions 9.0 References 10.0 Tables 11.0 Figures 12.0 Affected Documents 13.0 Attachment Sheet No.

1 2

3 4

5 5

5 1 8 24 32 43 43 45 50 72 78 78 Nuclear Common Revision 9 I INuclear Common Revision 9

I l CALCULATION CONTINUATION SHEET l SHEET 5 of 78 CALC. NO.: S-C-ZZ-MDC-1945

REFERENCE:

LCR S03-05 G. Patel/NUCORE, ORIGINATOR, DATE REV:

02/24/04 0

M Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/26/04

1.0 PURPOSE

The purpose of this calculation is to determine the post-Loss Of Coolant Accident (LOCA) doses at the Exclusion Area Boundary (EAB), Low Population Zone (LPZ), and Control Room (CR) using the Alternative Source Term (AST), guidance in the Regulatory Guide (RG) 1.183, RADTRAD3.02 computer code, and Total Effective Dose Equivalent (TEDE) dose criteria for the following post-LOCA release paths:

Containment Leakage Engineered Safety Feature (ESF) Leakage Containment Vacuum Relief Line Release Back-leakage to The Refueling Water Storage Tank (RWST) Leakage The CR dose is evaluated using a CR unfiltered inleakage of 150 cfm, which is greater than the nominal 100 cfm inleakage measured by the baseline Tracer Gas Testing (Ref. 9.33, Table 1).

2.0 BACKGROUND

The consequences of a LOCA were previously analyzed using the TID-14844 source term methodology to assess compliance with 10 CFR 50, Appendix A, General Design Criterion (GDC) 19 and 10 CFR 100 Section 100.11 dose criteria.

LCR S03-05 proposes to amend the SNGS Units 1 and 2 plant operating licenses to implement the full scope Alternative Source Term methodology in lieu of the TID-14844 source term methodology. The TEDE offsite dose acceptance criteria specified in Table 6 of Regulatory Guide 1.183 (Ref. 9.1) is implemented in lieu of the whole body and thyroid dose guidelines provided in 10 CFR 100.1 1. Also, the 5 rem TEDE control room dose acceptance criterion specified in 10 CFR 50.67 (Ref. 9.36) is implemented in lieu of the 5 rem whole body and equivalent organ dose guidelines provided in 10 CFR 50 Appendix A GDC 19.

The LOCA is analyzed using plant specific as-built design and licensing bases inputs which are compatible to the TEDE dose criteria. The LOCA analysis is performed using the guidance in Regulatory Guide 1.183 and its Appendix A (Ref. 9.1).

3.0 ANALYTICAL APPROACH:

3.1 Core Inventory The Salem I & 2 existing core inventory consists solely of noble gas and halogen isotopic inventories (Ref. 9.3, Table 2). The AST analysis requires radionuclide inventories for the additional radionuclide groups shown in Table 5 of Regulatory Guide 1.183 (Ref. 9.1). Therefore, the Salem 1 & 2 post-LOCA core inventory required for the AST analysis is developed. The existing iodine and noble gas core inventory is based on a core thermal power level of 3,600 MWt (Ref. 9.3, Table 1). The Salem current I Nuclear Common Revision 9 l I Nuclear Common Revision 9

I CALCULATION CONTINUATION SHEET I

SHEET 6 of 78 CALC. NO.: S-C-ZZ-MDC-1945

REFERENCE:

LCR S03-05 G. Patel/NUCORE, ORIGINATOR, DATE REV:

02/24/04 0

l M Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02126/04 licensed thermal power level of 3,459 MWt (Ref. 9.6.1) is multiplied by 1.05 to generate an AST core inventory at a thermal power level of 3,632 MWt (= 3,459 x 1.05), thereby providing a 2% margin for power measuring instrument uncertainty (Ref. 9.32) and additional margin for a future power uprate.

The increased thermal power level of 3,632 MWt is then divided by 3,600 MWt to obtain an adjustment factor of 1.009 (= 3,632 MWt/3,600 MWt) that is applied to the existing iodine and noble gas activity to obtain the normalized core inventory for 3,632 MWt (Table 1). The comparison of the RADTRAD default nuclide inventory file (NIF) Pwr def.nif and the Westinghouse plant-specific core iodine and noble gas inventory is shown in Table IA, and indicates that the plant-specific iodine inventory is approximately 10% higher than the RADTRAD default inventory. The normalized iodine and noble gas from the Table 1 inventories are divided by the core thermal power level of 3,632 MWt to develop iodine and noble gas inventories suitable for input into the RADTRAD Nuclide Inventory File (NIF)

(Table 2). The aerosol nuclide inventories (Ci/MWO) are obtained from the RADTRAD3.02 (Ref. 9.2) default nuclide inventory file (NIF) Pwr def.nif and listed in Table 3 with the iodine and noble gas inventories obtained from Table 2. Table 3 provides the NIF required for the Salem AST analysis. The RADTRAD default NIF Pwr_def.nif is modified using the information developed in Table 3. The newly developed NIF Salemdef.nif is used with the thermal power level of 3,632 MWt in the containment leakage analyses. Since the SNGS plant-specific inventory is 10% higher than the RADTRAD default value, the containment leakage and CR filter shine doses that are calculated with the RADTRAD aerosol inventory are increased by 10% (Sections 6.11 & 6.13). The ESF leakage dose is entirely from the iodine inventory and the external cloud dose is a whole body dose, both of which are calculated with the SNGS plant-specific inventory in Table I and therefore need not be increased.

3.2 Containment Leakage 3.2.1 Activity Transport in Primary Containment The Salem containment spray covers 75% of the containment volume (Ref. 9.34, Section 3.4.1.4). Five containment fan cooling units (CFCUs) are to be operable per Technical Specification 3.6.2.3 (Ref. 9.6.7). The ventilation P&IDs (Ref. 9.47) and the ventilation duct location drawings (Ref. 9.48) show that the CFCUs take suction from the area above the operating floor (EL 130'-0"), and discharge into a ring header located just below the operating floor. The ventilation ring header distributes air to both the sprayed and unsprayed regions in the containment (Refs. 9.47 and 9.48). The locations of the CFCUs are shown in References 9.28.d through 9.28.f. During normal operation, the 5 CFCUs deliver 550,000 cfm to the various containment areas (Ref. 9.47). Of the 550,000 cfm of normal flow, approximately 185,000 cfm is distributed above the operating floor and 60,000 cfm is distributed to the steam generator cubicles (Refs. 9.47, 9.48, & 9.19). Therefore, a total of 245,000 cfm out of the 550,000 cfm of CFCUs air is distributed from the unsprayed to sprayed regions, which is 44.55% of the total CFCUs flow (245,000/550,000 x 100% = 44.55%).

During an accident each CFCU is designed to supply a nominal flow of 39,000 cfm (Ref 9.14, Section 2.2.d). Only two CFCUs are conservatively assumed to be available during a LOCA, which provide a total air circulation flow of 78,000 cfm (2 x 39,000 cfm = 78,000 cfm). Assuming that the post-LOCA distribution of air to the various areas remains the same as that in the normal operating condition, the I Nuclear Common Revision 9 1 Nuclear Common Revision 9

CALCULATION CONTINUATION SHEET SHEET 7 or 78 CALC. NO.: S-C-ZZ-MDC-1945

REFERENCE:

LCR S03-05 G. Patel/NUCORE, ORIGINATOR, DATE REV:

02/24/04 0

M Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/26/04 operation of two CFCUs will distribute 55.45% of the total air circulation flow of 78,000 cfm, which is 43,251 cfm, from the sprayed region to the unsprayed regions.

Per RG 1.183, Appendix A, Section 3.3 (Ref. 9.1), the mixing rate attributed to natural convection between sprayed and unsprayed regions of the containment building, provided that adequate flow exists between these regions, is assumed to be two turnovers of the unsprayed regions per hour. The post-LOCA two CFCU air flow of 43,251 cfm from the sprayed to unsprayed regions satisfies the RG 1.183 requirement. Therefore, the mixing rate between the sprayed and unsprayed regions is assumed to be 2 turnovers of the unsprayed region per hour, which is 21,833 cfm [(2/hr x 0.25 x 2.62E+06 ft3) / (60 minutes/hr) = 21,833 cfm]. This mixing rate is conservatively less than the post-LOCA two CFCU air flow of 43,251 cfm from the sprayed to unsprayed regions.

3.2.2 Extended Containment Spray Operation Containment spray removal of iodine and particulates (aerosols) is assumed to be initiated at 90 seconds after the start of the LOCA event (this is conservatively later than the 85 second initiation time specified in Reference 9.12, page 95). In the initial containment spray injection phase the spray water is drawn from the refueling water storage tank (RWST). The injection phase terminates at 48 minutes after the start of the LOCA (Ref. 9.11), at which time the containment spray recirculation phase begins and the spray water is drawn from the containment sump. A containment spray interruption of 5 minutes is expected during the transition from the injection to the recirculation phase. This containment spray interruption is conservatively assumed to be 10 minutes in this analysis (from 48 to 58 minutes). During the interruption, the containment spray is not credited in removing the aerosol and elemental iodine activities from the containment atmosphere. The containment spray injection phase elemental removal coefficient is calculated to be 29 he' in Reference 9.8, Table 4-2 using the Salem plant-specific design condition parameters. However, Standard Review Plan 6.5.2 (Ref. 9.9, page 6.5.2-11) limits the use of the elemental iodine removal coefficient XE to 20 hl". Therefore, a XE of 20 hr'l is used during the injection phase. The spray aerosol removal coefficient Xp in the injection phase is calculated to be 4.44 hl' in Section 6.2 using the plant-specific design input information. During the injection phase, one containment spray pump (CSP) is assumed to operate (postulating a single failure of the other pump) at a minimum flow rate of 2,600 gpm (Ref. 9.8, Table 4-2). The containment spray flow rate is reduced to 1,900 gpm during the recirculation phase (Ref. 9.34, Section 4.0). Therefore, for the recirculation phase XE and Xp are re-calculated in Section 6.3 based on the reduced flow rate.

SRP 6.5.2 (Ref. 9.9, Section 4.c.(4).d) sets forth a maximum decontamination factor (DF) of 200 for elemental iodine based on the maximum iodine activity in the primary containment atmosphere when the sprays actuate, divided by the activity of iodine remaining at some time after decontamination.

Regulatory Guide 1.183 specifies that the maximum activity to be used in determining the containment spray DF is defined as the iodine activity in the columns labeled "Total" in Table 2 of RG 1.183 (Ref. 9.1, Appendix A, Section 3.3) multiplied by 0.05 for elemental iodine and by 0.95 for particulate iodine (i.e., aerosols are treated as particulates in SRP methodology). The DF for the containment atmosphere achieved by the containment spray system can be determined using the following equation (Ref. 9.9, page 6.5.2-12):

Nuclear Common Revision 9

WI-I CALCULATION CONTINUATION SHEET ISHEET 8 of 78 CALC. NO.: S-C-ZZ-MDC-1945

REFERENCE:

LCR S03-05 G. Patel/NUCORE, ORIGINATOR, DATE IREV:

02/24/04 0

l l

M Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/26/04 DF =1+ V, H VI Where: V, = volume of liquid in containment sump Vc = containment net free volume less V.

H = partition coefficient This equation is used in Reference 9.15, page 4, to calculate the minimum DF of 546 for the Salem Fuel Upgrade/Margin Recovery (FU/MR) project. The DF of 100 (which is smaller than both 200 and 546) is conservatively used in the containment leakage analysis.

The SRP 6.5.2 also states that the particulate iodine removal rate should be reduced by a factor of 10 when a particulate DF of 50 is reached (Ref. 9.9, Section 4.c.(4)). The maximum iodine activity in the containment atmosphere occurs at the end of early-in-vessel release phase, which is 1.8 hrs after a LOCA (Ref. 9.1, Table 4). Therefore, the corresponding XE and Xp cutoff times are calculated in Section 6.4 based on an assumption that the maximum iodine activity occurs at 1.8 hrs after onset of a LOCA. Although one containment spray can be operable for a long time after a LOCA, containment spray operation is assumed to be terminated at 4.0 hrs after a LOCA. The containment leakage model is shown in Figure 1. The corresponding assumptions and design inputs are shown in Sections 4.0 and 5.0.

The resulting containment leakage doses are summed with the doses from other post-LOCA sources in Section 7.0.

3.2.3 Long-Term Iodine Partition RG 1.183, Appendix A, Section 1, requires evaluation of the re-evolution of iodine for a sump pH value of less than 7. During the injection and recirculation phases, the sump water pH will remain at > 7 (Ref. 9.37, Section 5.0) including the effect of acids and bases created during the LOCA event and radiolysis products. Consequently, the re-evolution of dissolved iodine from the sump is not considered.

3.3 Engineered Safety Feature (ESF) System Leakage ESF systems that recirculate containment sump 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 in the Residual Heat Removal (RHR), Safety Injection (SI), Containment Spray (CS) and Chemical & Volume Control (CVC) systems (Ref. 9.16, Attachment 8.1). The radiological consequences from the postulated leakage are analyzed and combined with consequences postulated for other fission product release paths to determine the total calculated radiological consequences from the LOCA. The recirculation system components subject to leak during the recirculation phase are located in the Auxiliary Building (AB) at EL 84'-0" and 45'-0" (Refs. 9.28.b through 9.28.e, and Ref. 9.30). Back-leakage to the Refueling Water Storage Tank (RWST) is discussed in Section 3.5.

Nuclear Common Revision 9 I uclear Common Revision 9 l

I CALCULATION CONTINUATION SHEET SHEET 9 of 78 CALC. NO.: S-C-ZZ-MDC-1945

REFERENCE:

LCR S03-05 G. Patel/NUCORE, ORIGINATOR, DATE REV:

02/24/04 0

M Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/26/04 3.3.1 Post-LOCA Iodine Source Term In Sump Water Regulatory Guide 1.183, Appendix A, Section 5.1, requires that with the exception of noble gases, all the fission products released from the fuel to the containment (as defined in Table 2 of Ref. 9.1) should be assumed to instantaneously and homogeneously mix in the primary containment sump water. In lieu of this deterministic approach, suitably conservative mechanistic models for the transport of airborne activity in containment to the sump water may be used. The sump water activity in this analysis is established in the following section based on a suitably conservative iodine transport of containment airborne activity to the sump, which is permitted in RG 1.183, Appendix A, Section 5.1. The requirements of Section 5.1 of RG 1.183, Appendix A for a suitably conservative iodine transport model is adopted in this AST dose analysis and justified in the following section:

Forty percent of iodine released from the fuel to the containment is assumed to be transported to the containment sump (Ref. 9. 1, Table 2). If the sump pH is controlled at values 7 or greater, the radioiodine released from the fuel to containment should be assumed to be 95 percent cesium iodide (CsI),

4.85 percent elemental iodine, and 0.15 percent organic iodide (Ref. 9.1, Section 3.5 and Appendix A, Section 2). The containment sump pH is maintained at greater than 7 during and following a LOCA at the Salem plants (Ref. 9.37, page 4). With the exception of elemental and organic iodine and noble gases, fission products should be assumed to be in particulate form (Ref. 9.1, Section 3.5). Based on these regulatory requirements, the chemical form of sump water iodine could be 4.85% elemental iodine, which is volatile and subject to become airborne from the ESF leakage in proportion to a flashing rate, and 95% cesium iodide (CsI), which remains waterborne provided the pH of the containment sump water is > 7. If the sump water pH is less than 7, the iodine will be re-evolved as indicated by the discussion in Section 1.3 of Appendix B of Reference 9.1. The remaining 0.15% organic iodide is in gaseous form, is released to the containment atmosphere, and has its dose contribution accounted for in the containment leakage.

3.3.2 Post-LOCA Iodine Transport To Sump Water Analysis The operation of the containment spray system is extended to 4.0 hrs following a LOCA. The quantity of elemental iodine transported into the sump water is calculated using conservative design inputs and RADTRAD 3.02 code. Although 75% of the containment volume is sprayed and 25% of the containment volume is unsprayed, the containment is assumed to be a well mixed single volume to maximize the iodine transport to the containment sump.

Many of the parameters that make spray and deposition models conservative with regard to a containment airborne leakage analysis are non-conservative with regard to an analysis of the buildup of sump iodine activity. A containment leakage analysis conservatively maximizes the release of activity to the environment by maximizing the containment airborne activity concentrations. These containment leakage analysis conservatisms include consideration of a single failure of one CS pump, a smaller elemental iodine removal coefficient (use of regulatory value per SRP 6.5.2 versus a higher calculated value), and a lower DF for the spray cutoff time for elemental iodine removal. This containment leakage analysis modeling is non-conservative when calculating the buildup of sump water iodine activity.

I Nuclear Common Revision 9 I

I CALCULATION CONTINUATION SHEET SHEET IO of 78 CALC. NO.: S-C-ZZ-MDC-1945

REFERENCE:

LCR S03-05 G. Patel/NUCORE, ORIGINATOR, DATE IREV:

02/24/04 0

l M Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/26/04 Therefore, the following conservative design inputs are used to maximize the iodine transport to the sump water:

1.

The operation of two CS pumps is assumed for 60 minutes (Ref. 9.8, Section 4.1) instead of the one CS pump for 48 minutes used in the containment leakage analysis (DI 5.3.2.16).

2.

The total elemental iodine removal coefficient of 58 hri (2 x 29 hr 4 = 58 hr 1) (Ref. 9.8, Table 4-2) is used instead of the 20 hr-1 used in the containment leakage analysis during the injection phase (DI 5.3.2.12).

3.

A larger DF of 200 is used for the elemental iodine removal coefficient to increase the elemental iodine removal cutoff time (DI 5.3.2.15).

4.

A single well mixed containment volume is used instead of dividing the containment volume into two regions - 75% sprayed volumes and 25% unsprayed volumes.

5.

The containment spray interruption is not credited to maximize the iodine transport to the sump The RADTRAD computer code (Ref. 9.2) precisely balances the post-LOCA airborne iodine activity released from the core by distributing it in three different regions: containment, sump, and environment.

The iodine transport to sump model is developed using the above conservative design input to maximize the iodine activity in the sump water. The relation between the elemental iodine activity and atom (atom/curie) is established in Table 9 using the iodine nuclide information obtained from iodine transport to sump RADTRAD output file SI50CS9OCLO.oO. The isotopic fractions of elemental iodine are calculated in Table 9. The RADTRAD code calculates the cumulative elemental iodine atoms in the sump at the end of each time interval. The containment spray operation is assumed to be terminated at 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after it is initiated. The cumulative elemental iodine atoms of 9.4225E+22 introduced into the containment sump via spray washout is obtained from S150CS90CLO1.o0 file at the end of 4 hrs, which is converted into the elemental iodine isotopic activities in Table 10 using the isotopic fractions and atom/curie information established in Table 9. The I-13 1 elemental sump iodine activity of 1.939E+06 Ci (Table 11) is compared with 4.85% elemental iodine activity of the expected 40% sump iodine activity of 1.938E+06 Ci. The comparison shows that the calculated elemental iodine activity in the sump is equivalent to the expected sump elemental iodine (1.939E+06/1.938E+06 x 100% = 100%).

3.3.3 ESF Leakage Release Path The ESF leakage during the post-LOCA injection and recirculation phases includes leakage through valve packing glands; pump shaft seals, flanged connections, and other similar components associated with the containment spray pumps, charging pumps, safety injection pumps, and residual heat removal (RHR) pumps. These release points are located in the auxiliary building (Refs. 9.28 & 9.30). These components are located in various cubicles at different elevations in the auxiliary building with appropriate floor slopes that drain any leakage into two sumps located in the auxiliary building basement (Ref. 9.31). Per dimensions provided in References 9.30.a and 9.31, the total sump capacity is 387 ft3

(= 2 sumps x 3' wide x 9' long x 7.167' deep [= 45'-2"- 38']). This total sump capacity is not adequate to hold the total ESF leakage of 5,775 ft3 (= 1 gpm x 60 min/hr x 720 hr x (7.481 gal/ft 3)') modeled in this AST dose analysis. Therefore, the basement floor of the RHR pump cubicles will be flooded with a pool of ESF water. The RHR pump rooms can be accessed from the upper floors via an enclosed Io Nuclear Common

~Revision 9 I

I Nuclear Common Revision 9 I

I CALCULATION CONTINUATION SHEET ISHEET 11 of 78 CALC. NO.: S-C-ZZ-MDC-1945

REFERENCE:

LCR S03-05 G. Patel/NUCORE, ORIGINATOR, DATE IREV:

02/24104 0

l l

M Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/26104 staircase with a closed door (either Unit 1 Stair #3 or Unit 2 Stair #4 - Refs. 9.30.a & 9.40), which confines the airborne ESF leakage iodine activity to the RHR pump rooms. The locations of various pumps are either in the middle or the basement of the auxiliary building, therefore, ESF leakage releases to the outside environment through building cracks and openings are highly improbable due to the existence of multiple barriers (walls, ceilings, floors) between the ESF leakage radiation sources and the environment (Refs. 9.28 & 9.31). The 1 gpm ESF leakage rate is insufficient to pressurize the auxiliary building because the pump rooms are equipped with dedicated fan coolers to remove the heat dissipated from the pump operations and ESF leakage (Ref. 9.29). The mixing of ESF leakage in the auxiliary building air spaces and dilution with other exhaust air flows in the plant vent is not credited in this analysis. A discussion of back-leakage to the RWST is provided in Section 3.5.

3.3.4 Multiple Barriers To Confine CsI Within Auxiliary Building Sump pH The containment sump water pH will be maintained > 7 during and following a LOCA (Ref. 9.37, Section 5.0). The auxiliary building ESF recirculation pump rooms floors are painted with a surface coat of 10 mil dry thickness of Phenoline 300 surfacer (Ref. 9.38, pages 3 and 5). The walls & ceilings are painted with 10 mil wet film thickness of Carboline 195/Carboline 300 surfacer (Ref. 9.38, pages 3 and 6). The pH of the concrete surfaces is maintained between 7 and 8 before the surfacer is applied (Ref. 9.38, page 1). The decontamination surfacer coating is a barrier that prevents the chemical reaction of ESF leakage (borated water) with the concrete surfaces, which are already prepared to have a pH greater than 7 before the surfacer is applied.

The aerosols from limestone concrete contain the basic oxides CaO, Na2O, and K20 (Ref. 9.39, Section 2.3.2). The results of a series of pH tests that were run at ORNL on aerosol material indicate that the concrete-core reaction is basic and it has a tendency to increase the pH (Ref. 9.39, Table 2.4).

Assuming the floor surface coating is damaged in the RHR pump rooms, where the ESF leakage from various safety pumps is drained and colleted on the floor, the chemical reaction of the borated ESF leakage water with the concrete under the damaged floor surface coating will also be basic. Therefore, the reaction will increase the pH of the ESF leakage and provide a stronger bond for the dissolved aerosol, which prevents them from becoming airborne. This justifies that only the elemental iodine species in the ESF leakage become airborne and that use of the conservatively calculated sump elemental iodine fraction in Table 11 is conservative and appropriate.

Aerosol Removal By Pump Room Cooler Roughing Filters The air supply and exhaust from auxiliary building pump rooms are shown in Reference 9.29. Pump rooms are equipped with dedicated air cooler units with commercial grade roughing filters (Ref. 9.29),

which are capable of removing particulates from the recirculated air. The containment spray, charging, safety injection, and RHR pump rooms have small outside air supplies from the auxiliary building ventilation system (ABVS) compared to their large recirculation flows through the room coolers (Refs.

9.17 & 9.29).

Nuclear Common Revision 9

I I

CALCULATION CONTINUATION SHEET SHEET 12 of 78 CALC. NO.: S-C-ZZ-MDC-1945

REFERENCE:

LCR S03-05 G. Patel/NUCORE, ORIGINATOR, DATE REV:

02/24/04 0

l M Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/26104 If any dissolved aerosol becomes airborne, especially the CsI displaced mechanically (i.e., CsI release in the aerosol form without its chemical conversion into volatile iodine species) from the ESF leakage by thermal-dynamic force or its entrainment in the flashed ESF leakage, a large portion of the airborne CsI would be captured on the room cooler roughing filters depending on the filter aerosol removal efficiencies.

Aerosol Removal By Gravitational Deposition The pump room walls and ceilings, pumps and equipment surfaces, fan coolers, and HVAC ducts provide a very large surface area for gravitational deposition of the airborne aerosols (CsI). Any airborne aerosol that would escape from the pump rooms by being entrained with the pump room exhaust air, and subsequently released to the environment via the plant vent, will be gravitationally removed from the air during its migration to receptor locations. Consequentially, the cesium iodide, which mechanically becomes airborne, will be removed by gravitational deposition.

If there would be mechanically separated CsI from the sump water by its entrainment in the flashed liquid, it would be removed from the air by the roughing filters associated in the room coolers and gravitational deposition on various surfaces.

3.3.5 ESF Leakage Dose Analysis ESF system leakage is conservatively assumed to begin at 20 minutes, representing the earliest possible recirculation start time with the ECCS operating at maximum capacity (Ref. 9.8, page 4-1). The design basis ESF leakage rate of 3,790 cc/hr (= I galfhr) is increased to I gpm. Per the guidance in Regulatory Guide 1.183, the ESF leakage of I gpm is doubled and an iodine flashing fraction of 10 percent is applied (Ref. 9.1, Appendix A, Sections 5.2 and 5.5), resulting in an equivalent ESF leakage rate to the environment of 0.2 gpm (= 0.02673 cfm).

The auxiliary building ventilation system (ABVS) charcoal filter is not credited in the AST ESF leakage analysis.

Based on the discussion in Sections 3.3.2 and 3.3.4 above, it is determined that the 4.85% of elemental iodine released from the core to containment atmosphere and transported to the containment sump becomes the iodine source for the ESF leakage. The elemental iodine of 4.85% is increased by a factor of 3 to add a large conservatism in the resulting ESF leakage doses. Table 11 shows that the resulting elemental iodine represents a 0.05824 fraction of the core iodine inventory. Therefore, the RADTRAD model is developed for the ESF leakage analysis using the RADTRAD release fraction & timing (RFT) file with only an iodine release fraction of 0.05824 and applicable design inputs and assumptions in Sections 4.0 & 5.0. The sump water volume of 43,930 ft is conservatively used based on the water inventory at 51 minutes after a LOCA (Ref. 9.15), which is less than the sump water volume of 45,510 ft3 used in the sump pH analysis (Ref. 9.37, Section 4.3.4). The resulting doses are summed with the doses from other post-LOCA sources in Section 7.0.

I Nuclear Common Revision 9 1 Nuclear Common Revision 9 I

A I

CALCULATION CONTINUATION SHEET ISHEET 13 of 78 CALC. NO.: S-C-ZZ-MDC-1945

REFERENCE:

LCR S03-05 G. Patel/NUCORE, ORIGINATOR, DATE REV:

02/24/04 0

M Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/26/04 3.4 Containment Vacuum Relief Line Release The containment vacuum relief line release (VRLR) occurs following the Large Break LOCA, and before containment isolation. The entire reactor coolant system (RCS) inventory is assumed to be instantaneously released and homogeneously mixed in the containment atmosphere. The containment pressurization due to the RCS mass and energy release combined with the presence of the containment vacuum relief line result in the potential for a limited release of airborne activity to the environment.

100% of the radionuclide inventory in the RCS liquid is assumed to be released into the containment at the initiation of the LOCA (Ref. 9.1, Appendix A, Section 3.8). A release of gap activity into the containment is not considered since the containment vacuum relief line release duration of 5 seconds terminates this release path prior to the onset of the gap release phase (Ref. 9.1, Table 4).

The RCS activity inventory is based on the technical specification for RCS equilibrium activity (Ref. 9.6.12). Iodine spikes are not considered (Ref. 9.1, Appendix A, Section 3.8). Per Technical Specification (TS) LCO for Specific Activity (Refs. 9.6.12), the primary coolant equilibrium iodine concentration permitted by the technical specifications is 1 piCi/gm Dose Equivalent (DE) 1-13 1.

Technical Specification Section 1.10 (Ref. 9.6.13) defines DE 1-131 as that concentration of 1-131 (in p.Ci/g) which alone would produce the same thyroid dose as the quantity and isotopic mixture of 1-131, 1-132,1-133, I-134, and I-135 actually present using the thyroid dose conversion factors (DCFs) specified in Table III of TID-14844; the TS Section 1.10 definition of DE 1-131 will be revised to be based on the DCFs specified in Federal Guidance Report 11 (Ref. 9.27). Therefore, this analysis calculates DE-131 in terms of the thyroid dose conversion factors specified in FGR 11.

The plant-specific RCS iodine concentrations corresponding to 1% fuel defects are obtained from Reference 9.3 (Table 4), and listed in Table 5. The iodine dose conversion factors are developed in Table 4 using the information in Reference 9.27 and used in Table 5 to establish an iodine scaling factor based on the iodine concentration of I pCi/g DE 1-131, which is then used in Table 6 to convert the iodine concentrations of 1% fuel defects to 1 jiCi/g DE I-13 1. The total isotopic iodine activity in the RCS is conservatively calculated in Table 7 using the RCS mass determined in Section 6.7. Table 5 calculates a 1 percent failed fuel primary coolant iodine concentration of 3.58 jiCi/gm DE I-13 1, which is conservatively greater than the TS limit of 1.0 IiCilgm DE 1-131. Since 1 percent failed fuel introduces more iodine activity into the primary coolant than is allowed by the Tech Specs, it can be conservatively assumed that 1 percent failed fuel introduces more non-iodine activity into the RCS than is allowed by the Tech Spec limit of 10 O/E-bar. Therefore, the RCS noble gas concentrations corresponding to 1% fuel defects are conservatively used in Table 7 to determine the total RCS noble gas activity released in the containment during the VRLR based on the total coolant mass in the RCS.

The isotopic activity is divided by the core thermal power level of 3,632 MWt in Table 8 to obtain the isotopic Ci/MWt, which is used to develop the RADTRAD NIF SPURGE def.txt. The RADTRAD dose conversion file (DCF) Fgrl 0&11.inp is modified to include short lived noble gas isotopes Kr-83m, Xe-131m, Xe-133m, Xe-135m, Xe-138 in the Westinghouse core inventory (See Table 1) The newly developed SPURGEdef.txt and SPURGEFGRI 1&12.txt files are used to calculate the CR and site boundary doses due to the VRLR. The CR is assumed to be in a normal mode operation during the Nuclear Common Revision 9 I Nuclear Common Revision 9 1

CALCULATION CONTINUATION SHEET SHEET 14 of 78 CALC. NO.: S-C-ZZ-MDC-1945

REFERENCE:

LCR S03-05 G. PateUNUCORE, ORIGINATOR, DATE REV:

02t24/04 0

M Drucker/NUCORE, REVIEWERIVERIFIER, DATE 02t26/04 VRLR. The applicable assumptions and design inputs are shown in Sections 4.0 and 5.0. The resulting doses are summed with the doses from other post-LOCA sources in Section 7.0.

3.5 Post-LOCA Back-leakage To The Refueling Water Storage Tank (RWST)

The Salem I & 2 RWSTs are located west of the containment buildings (Refs. 9.21.c and 9.28.a). The relative locations of plant vents, RWSTs, and CR air intakes are shown in Figure 4. The shortest distance of 76.74 meter [((23.5)2 + (250.62')2 y"2 = 251.72 ft = 251.72 ft/(3.28 ft/n) = 76.74 m]

between the centerlines of the Unit 2 RWST and the Unit 2 CR air intake is more than double the distance of 30.25 m [((18.38 )2 + (97.5 )2).In = 99.22 ft = (99.22 ft)/(3.28 ft/in) = 30.25 m] between the centerlines of the plant vent and the CR air intake (Figure 4). The X/Qs for the RWST release would be proportionately lower than those for the containment and ESF leakage releases via the plant vent. The worst case back-leakage to the RWST is estimated to be 100 cc/hr (Ref. 9.13, page 3) equal to 5.886E-05 cfhi (100 cc/hr x 1/60 hr/min x 1/(3785 cc/gal) x 1/(7.481 gal/fl3) = 5.886E-05 cfm), which is negligibly small. The negligibly small RWST back-leakage is further diluted into a large RWST air space volume of 35,806 ft3 before it is vented to the atmosphere. The post-LOCA back-leakage to the RWST will result in insignificantly small dose consequences, therefore, its dose contribution is not analyzed.

3.6 Control Room Dose Consequences The TEDE analysis considered the radioactive releases from the containment & ESF leakages and containment vacuum relief line release for exposure to control room personnel. The following radioactive sources are analyzed for CR operator exposure:

I.

Contamination of the control room atmosphere by the intake or infiltration of the radioactive material contained in the radioactive plume released from the facility, 2

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, 3

Radiation shine from the external radioactive plume released from the facility, 4

Radiation shine from radioactive material in the reactor containment, 5

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.

3.6.1 & 3.6.2 CR Airborne Doses From Filtered and Unfiltered Inleakages Since the radioactive material in the areas and structures adjacent to the CR envelope is the same as that in the post-LOCA radioactive plume released from the facility, Items 3.6.1 & 3.6.2 collectively represent the CR airborne doses from the filtered intake and unfiltered inleakage from the various post-LOCA sources which are shown in Figure 3. The CR airborne doses from these various post-LOCA l Nuclear Common Revision 9 Nuclear Common Revision 9 I

A CALCULATION CONTINUATION SHEET ISHEET 15 of 78 CALC. NO.: S-C-ZZ-MDC-1945

REFERENCE:

LCR S03-05 G. Patel/NUCORE, ORIGINATOR, DATE REV:

02t24/04 0

M Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/26/04 release paths are discussed in Sections 3.2 through 3.4 and the resulting doses are summarized in Section 7.0.

3.6.3 Radiation Shine From External Radioactive Plume The post-LOCA radioactive plume released from the plant vent contains the radioactive sources from the containment & ESF leakages and the containment vacuum relief line release. The CR envelope having two feet of concrete bulk shielding (Refs. 9.22 & 9.23) is submerged within the post-LOCA radioactive plume and the CR operator is exposed to gamma dose through the walls and roof. The RADTRAD3.02 code calculates the site boundary whole body gamma dose based on the semi-infinite cloud immersion (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 files SI5OCS75CLOl.psf and S75ESFIGPM.psf are modified by replacing them with the newly developed x/Qs for the CR center location. The semi-infinite gamma dose is calculated at the center of CR at the roof elevation. The ARCON96 code (Refs. 9.43 & 9.44) is used to calculate the new set of x/Qs for the CR center location using the parameters calculated in Section 6.8. ARCON96 source/receptor geometry is shown in Figure 5 with the essential input parameters. The remaining ARCON96 code related modeling techniques are similar to those discussed in Reference 9.5. The new set of x/Qs are as follows:

Time Unit 1 Plant Vent to Interval Unit I CR Center (hr)

A 0-2 5.63E1-04 2-8 3.99E-04 8-24 1.78E-04 24-96 1.3 1E-04 96-720 1.09E-04 A From ARCON96 Run CONTROL.log The RADTRAD3.02 files SEMIC75CLO.psf for the containment leakage and SEMI75ESF0O.psf for the ESF leakage are developed using the newly developed x/Qs to calculate the semi-infinite cloud gamma dose at the center of the CR. The containment vacuum relief dose contribution is extremely small (Section 7.0), therefore, it is excluded from this analysis. The total whole body gamma dose is calculated to be 4.33 rem (2.89 rem from containment leakage + 1.44 rem from ESF leakage). This is a semi-infinite dose above the center of the CR at the roof EL 140'-0" (Ref. 9.23.c). This gamma dose is attenuated by the 2-foot thick concrete roof above the CR (Ref. 9.23.c) and an air space of 10 feet (140'0" - 2'0" - 122'0" - 6'0" = 10'0") assuming a 6-foot tall operator standing on the CR floor at EL 122'-0". The energy dependent gamma dose attenuation factors are calculated in Reference 9.42, Appendix III, and page 60. The gamma attenuation factors are substantially low for the low energy bins.

The average gamma dose attenuation factor of 0.0136 is conservatively used, which is the average for the five mean energies in the gamma energy range of 0.85 Mev through 2.75 Mev. The resulting gamma dose from the external cloud shine dose to a 6-foot tall CR operator standing on floor EL 122'-0" would I Nuclear Common Revision 9 1 Nuclear Common Revision 9

CALCULATION CONTINUATION SHEET SHEET 16 of 78 CALC.NO.: S-C-ZZ-MDC-1945

REFERENCE:

LCR S03-05 G. Patel/NUCORE, ORIGINATOR, DATE REV:

02/24/04 0

M Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/26/04 be 0.059 rem (4.33 rem x 0.0136 = 0.059 rem), which is added with the dose contribution from other post-LOCA sources in Section 7.0.

3.6.4 Radiation Shine From Radioactive Material In The Reactor Containment It is only the post-LOCA containment leakage activity confined in the containment dome air space above the operating floor that can conceivably contribute to the direct shine dose to the CR operator (lower locations within the containment building are heavily self-shielded by the reactor cavity and steam generator enclosure structures. The containment cylindrical concrete wall is 4'-6" thick below the spring line, and 3'-6" thick above the spring line (Ref 9.28). The CR outside concrete wall is 2'-0" thick (Ref. 9.23.a). The concrete shielding of at least 5'-6" between the post-LOCA radioactive source in the containment dome and the CR operating floor provides ample shielding to totally shield the post-LOCA containment shine dose to the CR operator.

3.6.5 Radiation Shine From CR Filter The CREACS charcoal filter trains are located west of the main CR (Refs. 9.21.a, 9.21.b, 9.22.a &

9.22.b). The plan and elevation views of the charcoal filters are shown in Reference 9.21. The charcoal filter geometry and orientation with respect to the combined CR is shown in Figure 6 based on References 9.20, 9.21 & 9.22. The floor is located at EL 122'-0" (Refs. 9.20.a and 9.21.a). The width (2'-6") and height (4'-3") of the charcoal bed are measured from References 9.21.a & 9.21.b. The post-LOCA aerosol buildup on the HEPA filter and the iodine buildup on the charcoal filter are calculated as follows:

3.6.5.1 Post-LOCA Iodine & Aerosol Activity On CR Charcoal/HEPA Filter-Containment Leakage The RADTRAD3.02 code calculates the cumulative elemental and organic iodine atoms and the aerosol mass deposited on the CR recirculation charcoal/HEPA filters. The CR intake filter iodine and aerosol activities are calculated in Section 6.9 for the containment leakage. The relationship between the aerosol mass and activity is established in Table 12 based on the information obtained from RADTRAD run SI 50CS75CL01.oO. The aerosol mass deposited on the CR HEPA recire filter is calculated by the RADTRAD code for the duration of the accident. Knowing the CR intake and recirc filtration flow rates, the relationship can be established to calculate the aerosol mass deposited on the intake HEPA filter as shown in Section 6.9.2. The total aerosol mass deposited on the CR HEPA filter due to the containment leakage is calculated in Section 6.9.2, which is used with the aerosol mass/activity relation established in Table 12 to calculate the aerosol isotopic activities deposited on the CR HEPA filter as shown Table 13.

The total (elemental + organic) iodine atoms deposited on the CR charcoal filter due to the containment leakage are calculated in Section 6.9.1. The iodine atoms are converted into the isotopic iodine activities in Table 14 using the atom/activity relation established in Table 9. The aerosol and iodine isotopic activities deposited on the CR charcoal/HEPA filter due to the containment leakage are shown in Table 15.

Nuclear Common Revision 9 I Nuclear Common Revision 9 l

I CALCULATION CONTINUATION SHEET lISHEET 17 of 78 CALC NO.: S-C-ZZ-MDC-1945

REFERENCE:

LCR S03-05 l

G. Patel/NUCORE, I

ORIGINATOR, DATE lREV:

02/24/04 0

l M Drucker/NUCORE, REVIEWER/ERIFIER, DATE 02/26/04 3.6.5.2 Post-LOCA Iodine & Aerosol Activity On CR Charcoal/HEPA Filter - ESF Leakage Similarly, the iodine and aerosol deposited on the CR charcoal/HEPA filter are calculated in Section 6.10 for the post-LOCA ESF leakage. The post-LOCA ESF leakage consists of a non-aerosol iodine release (97% of elemental iodine + 3% of organic iodine) (Ref. 9.1, Section 5.6) only, therefore, there is no aerosol mass deposited on the CR HEPA filter (S75ESF1 GPM.oO, CR Compartment Nuclide Inventory @ 720 hrs). The iodine deposited on the CR charcoal filter is calculated in Section 6.10.1. The iodine atoms are converted into the isotopic iodine activities in Table 16 using the atom/activity relation established in Table 9. The iodine isotopic activities deposited on the CR charcoal/HEPA filter due to the containment and ESF leakages are shown in Table 17.

3.6.5.3 MicroShield Analysis of CR Charcoal/HEPA Filter Shine The total charcoal/HEPA filter aerosol and iodine isotopic activities in Table 17 are input into the MicroShield (Ref. 9.45) Computer Run SCREAS.MS5 with the source geometry, dimension, and detector location as shown in Figure 6 to compute the direct dose rate from the CR filter. Due to the limitations of the MicroShield code, which calculates the dose rate at the dose point location within the projected area (width & height dimension), the dose point location at the center of the charcoal filter projected area is conservatively modeled because the actual dose point involves the larger direct distance in the air and slant distance in the concrete shielding. The 720-hrs direct dose from the CR filter shine is calculated in Section 6.11 using the CR occupancy factors and added to doses from other post-LOCA sources in Section 7.0.

3.7 Post-LOCA Activity Released To Environment The post-LOCA activity released to the environment for different time intervals are provided in Tables 18 through 22 for input into the Meteorological Information Data Acquisition System (MIDAS) computer code to assess the post-LOCA doses for emergency planning.

I Nuclear Common Revision 9 1 Nuclear Common Revision 9

I CALCULATION CONTINUATION SHEET ISHEET 18 of 78 CALC. NO.: S-C-ZZ-MDC-1945

REFERENCE:

LCR S03-05 G. Patel/NUCORE, ORIGINATOR, DATE REV:

02/24/04 0

M Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/26/04

4.0 ASSUMPTIONS

Regulatory Guide 1.183 (Ref. 9.1) provides guidance on modeling assumptions that are acceptable to the NRC staff for the evaluation of the radiological consequences of a LOCA. The following sections address the applicability of these modeling assumptions to this SNGS Units 1 and 2 LOCA analysis. These assumptions are incorporated as design inputs in Section 5.3 through 5.8 and are incorporated in this analysis.

4.1 Source Term Assumptions Acceptable assumptions regarding core inventory and the release of radionuclides from the fuel are provided in Regulatory Guide 1.183 (Reference 9.1, Sections 3.1 through 3.5) as follows:

4.2 Core Inventory The assumed inventory of fission products in the reactor core and available for release to the containment is based on the maximum power level of 3,632 MWt corresponding to current fuel enrichment and fuel burnup, which is 1.05 times the SNGS current licensed thermal power of 3,459 MWt (Ref. 9.6.1).

4.3 Release Fractions and Timing The core inventory release fractions, by radionuclide groups, for the gap release and early-in-vessel release for a Design Basis Accident (DBA) LOCA are listed in Design Input 5.3.1.5. These fractions are applied to the equilibrium core inventory described in Design Input 5.3.1.2 using the release timing specified in Design Input 5.3.1.4 (Ref. 9.1, Tables 2 & 4).

4.4 Radionuclide Composition The elements in each radionuclide group to be considered in the LOCA design basis analyses are shown in Design Input 5.3.1.3 (Ref. 9.1, Section 3.4 and Table 5).

4.5 Chemical Form The containment sump water pH is greater than 7 during and following a LOCA (Ref. 9.37, Section 5.0).

Consequently, the chemical forms of radioiodine released to the containment can be assumed to be 95 percent cesium iodide (CsI), 4.85 percent elemental iodine, and 0.15 percent organic iodine (Ref. 9.1, Appendix A, Section 2). These iodine chemical forms are shown in Design Input 5.3.1.6. With the exception of elemental and organic iodine and noble gases, fission products are assumed to be in particulate form (Ref. 9.1, Appendix A, Section 2).

4.6 Activity Transport in Primary Containment 4.6.1 The radioactivity released from the fuel is assumed to mix instantaneously and homogeneously throughout the free air volume of the primary containment (Ref. 9.1 Appendix A, Section 3.1).

I Nuclear Common Revision 9 l Nuclear Common Revision 9 I

I CALCULATION CONTINUATION SHEET ISHEET 19 of 78 CALC. NO.: S-C-ZZ-MDC-1945

REFERENCE:

LCR S03-05 G. Patel/NUCORE, ORIGINATOR, DATE REV:

02t24/04 0

M Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02t26/04 4.6.2 Reduction in airborne radioactivity in the containment by natural deposition within the containment is credited using the RADTRAD3.02 Powers model for aerosol removal coefficient with a 1 0-percentile probability (Ref. 9.1 Appendix A, Section 3.2 & Ref. 9.2 Section 2.2.2.1.2).

4.6.3 Reduction in airborne radioactivity in the containment by containment spray system is credited (Ref. 9.1, Appendix A, Section 3.3). The Salem plant-specific containment spray coverage analysis in Reference 9.34, Section 4.0 (using data presented in Reference 9.8, Table 4-1), indicates that the containment spray covers 75% of the containment volume. As discussed in Section 3.2.1, a mixing rate of 2 turnovers of the unsprayed region volumes per hour, which is equivalent to 21,833 cfm, is conservatively used in the containment leakage analysis.

4.6.4 The SRP 6.5.2 (Ref. 9.9) sets forth a maximum decontamination factor (DF) of 200 for elemental iodine based on the maximum iodine activity in the primary containment atmosphere when the sprays actuate, divided by the activity of iodine remaining at some time after decontamination. The maximum activity to be used in determining the DF is defined as the iodine activity in the columns labeled "Total" in Table 2 of RG 1.183 multiplied by 0.05 for elemental iodine and by 0.95 for particulate iodine. The minimum DF of 546 is calculated for the fuel upgrade/margin recovery project in Reference 9.15, page 4. The DF of 100 is conservatively used in the analysis for the containment leakage analysis.

4.6.5 The SRP 6.5.2 (Ref. 9.9) states that the particulate iodine removal rate should be reduced by a factor of 10 when a DF of 50 is reached. The cutoff times for the elemental and particulate removal coefficients are calculated in Section 6.4 based on an assumption that the maximum iodine activity occurs at 1.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after onset of a LOCA.

4.6.6 The primary containment is assumed to leak at the allowable Technical Specification peak pressure leak rate of 0.1% by weight for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (Ref. 9.1, Appendix A, Section 3.7, Ref. 9.6.4 & Ref. 9.7, Section 4.13). This leak rate is reduced to 0.05% by weight after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (Ref. 9.1 Appendix A, Section 3.7).

4.6.7 The containment vacuum relief release following the large break LOCA prior to containment isolation is analyzed using the 100% RCS inventory based on technical specification RCS equilibrium activity as discussed in Section 3.4 and resulting doses are summed with the postulated doses from the other release paths in Section 7.0 (Ref. 9.1, Appendix A, Section 3.8).

4.7 Assumptions on ESF System Leakage ESF systems that recirculate sump 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. This release source may also include leakage through valves isolating interfacing systems. The radiological consequences from the postulated leakage are analyzed and combined in Section 7.0 with consequences postulated for other fission product release paths to determine the total calculated radiological consequences from the LOCA. The following assumptions are acceptable for I Nuclear Common Revision 9 1 Nu l a Co m nR v s o

I I

CALCULATION CONTINUATION SHEET lISHEET 20 of 78 CALC. NO.: S-C-ZZ-MDC-1945

REFERENCE:

LCR S03-05 G. Patel/NUCORE, ORIGINATOR, DATE REV:

02/24/04 O

M Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/26/04 evaluating the consequences of leakage from ESF components outside the primary containment for a PWR, and are incorporated in the Design Inputs 5.4.1 through 5.4.8.

4.7.1 Regulatory Guide 1.183 (Appendix A, Section 5.1) states that with the exception of noble gases, all the fission products released from the fuel to the containment (as defined in Table 2 of Ref. 9.1) should be assumed to instantaneously and homogeneously mix in the primary containment sump water at the time of release from the core. Regulatory Guide 1.183 (Appendix A, Section 5.1) also states that in lieu of this deterministic approach, suitably conservative mechanistic models for the transport of airborne activity in containment to the sump water may be used. Note that many of the parameters that make spray and deposition models conservative with regard to containment airborne leakage are non-conservative with regard to the buildup of sump activity (See Section 3.3 for an evaluation of iodine transport to the containment sump).

4.7.2 The ESF leakage is taken as two times the sum of the simultaneous leakage from all components in the ESF recirculation systems above which the technical specifications, or licensee commitments to item III.D.1.1 of NUREG-0737, would require declaring such systems inoperable. The ESF leakage of 1 gpm is doubled (Section 3.3.5). The leakage should be assumed to start at the earliest time the recirculation flow occurs in these systems, which is 20 minutes, and end at the latest time the releases from these systems are terminated, which is 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> (Ref. 9.1 Appendix A, Section 5.2).

4.7.3 Consideration should also be given to design leakage through valves isolating ESF recirculation systems from tanks vented to atmosphere, e.g., emergency core cooling system (ECCS) pump miniflow return to the refueling water storage tank (Ref. 9.1 Appendix A, Section 5.2). For SNGS Units 1 and 2, the dose contribution from the RWST back-leakage is insignificantly small, therefore, it is not analyzed (Section 3.5).

4.7.4 With the exception of iodine, all radioactive materials in the recirculating liquid are assumed to be retained in the liquid phase (Ref. 9.1 Appendix A, Section 5.3).

4.7.5 Since the temperature of the ESF leakage exceeds 2127F, the fraction of total iodine in the liquid that becomes airborne is assumed equal to the fraction of the leakage that flashes to vapor. This flash fraction, FF, is determined using a constant enthalpy, h, process, to be no more than 10% based on the maximum time-dependent temperature of the sump water circulating outside the containment (Section 6.12). The FE of 10% is used for the entire duration of the ESF leakage (Ref. 9.1 Appendix A, Sections 5.4 and 5.5).

4.7.6 The radioiodine that is postulated to be available for release to the environment is assumed to be 97%

elemental and 3% organic. Reduction in release activity by dilution or holdup within buildings, or by ESF ventilation filtration systems, is not credited. The Auxiliary Building Ventilation System (ABVS) filtration is not credited. The ESF leakage is not assumed to be released to the environment with mixing with the auxiliary building volume (Ref. 9.1 Appendix A, Section 5.6).

4.8 Offsite Dose Consequences Regulatory Guide 1.183 (Ref. 9.1, Section 4.1) provides guidance to be used in determining the total effective dose equivalent (TEDE) for persons located at the exclusion area boundary (EAB) and at the NI l a ConR v s o I Nuclear Common Revision 9 1

CALCULATION CONTINUATION SHEET SHEET 21 of 78 CALC. NO.: S-C-ZZ-MDC-1945

REFERENCE:

LCR S03-05 1

lG. Patel/NUCORE, ORIGINATOR, DATE REV:

02/24/04 0

M Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/26/04 outer boundary of the low population zone (LPZ). The following sections address the applicability of this guidance to the SNGS Units 1 and 2 LOCA analysis. These assumptions are incorporated as design inputs in Section 5.6.1 through 5.6.6.

4.8.1 Consistent with RG 1.183 (Ref. 9.1, Section 4.1.1), this dose calculation determines the TEDE. The TEDE is the sum of the committed effective dose equivalent (CEDE) from inhalation and the deep dose equivalent (DDE) from external exposure. The calculation of these two components of the TEDE considers all radionuclides, including progeny from the decay of parent radionuclides that are significant with regard to dose consequences and the released radioactivity. These isotopes are listed in Section 5.3.1.2.

4.8.2 Consistent with RG 1.183 (Ref. 9.1, Section 4.1.2), the exposure-to-CEDE factors for inhalation of radioactive material are derived from the data provided in ICRP Publication 30, "Limits for Intakes of Radionuclides by Workers". This calculation models the CEDE dose conversion factors (DCFs) in the column headed "effective" yield doses in Table 2.1 of Federal Guidance Report 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion" (Ref. 9.27).

4.8.3 Consistent with RG 1.183 (Ref. 9.1, Section 4.1.3), for the first 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, the breathing rate of persons offsite is assumed to be 3.5 x 1 04 cubic meters per second. From 8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the accident, the breathing rate is assumed to be 1.8 x 104 cubic meters per second. After that and until the end of the accident, the rate is assumed to be 2.3 x 1 0 4 cubic meters per second. These offsite breathing rates are listed in Sections 5.6.2 and 5.6.4.

4.8.4 Consistent with RG 1.183 (Ref. 9.1, Section 4.1.4), the DDE is calculated assuming submergence in semi-infinite cloud assumptions with appropriate credit for attenuation by body tissue. The DDE is nominally equivalent to the effective dose equivalent (EDE) from external exposure if the whole body is irradiated uniformly. Since this is a reasonable assumption for submergence exposure situations, EDE is used in lieu of DDE in determining the contribution of external dose to the TEDE. This calculation models the EDE dose conversion factors in the column headed "effective" in Table III.1 of Federal Guidance Report 12, "External Exposure to Radionuclides in Air, Water, and Soil" (Ref. 9.41).

4.8.5 Consistent with RG 1.183 (Ref. 9.1, Sections 4.1.5 and 4.4), the TEDE is determined for the most limiting person at the EAB. The maximum EAB TEDE for any two-hour period following the start of the radioactivity release is determined and used in determining compliance with the dose criteria in 10 CFR 50.67 (Ref. 9.36). For the LOCA the postulated EAB doses should not exceed the criteria established in RG 1.183 Table 6. This assumption is incorporated as a design input in Section 5.6.5.

EAB Dose Acceptance Criterion:

25 Rem TEDE The RADTRAD Code (Refs. 9.2 & 9.4) used in this analysis determines the maximum two-hour TEDE by calculating the postulated dose for a series of small time increments and performing a "sliding" sum over the increments for successive two-hour periods. The time increments appropriately reflect the I Nuclear Common Revision 9 I Nuclear Common Revision 9 I

I CALCULATION CONTINUATION SHEET ISHEET 22 of 78 CALC. NO.: S-C-ZZ-MDC-1945

REFERENCE:

LCR S03-05 G. Patel/NUCORE, ORIGINATOR, DATE REV:

02/24/04 0

l M Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/26/04 progression of the accident to capture the peak dose interval between the start of the event and the end of radioactivity release.

4.8.6 Consistent with RG 1.183 (Ref. 9.1, Sections 4.1.6 and 4.4), the 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. 9.36). For the LOCA the postulated LPZ doses should not exceed the criteria established in RG 1.183 Table 6. This assumption is incorporated as a design input in Section 5.6.6.

LPZ Dose Acceptance Criterion:

25 Rem TEDE 4.8.7 Consistent with RG 1.183 (Ref. 9.1, Section 4.1.7), no correction is made for depletion of the effluent plume by deposition on the ground.

4.9 Control Room Dose Consequences Regulatory Guide 1.183 (Ref. 9.1, Section 4.2) provides guidance to be used in determining the total effective dose equivalent (TEDE) for persons located in the control room (CR). The following sections address the applicability of this guidance to the SNGS Units 1 and 2 LOCA analysis. These assumptions are incorporated as design inputs in Sections 5.5.1 through 5.5.14.

4.9.1 Consistent with RG 1.183 (Ref. 9.1, Section 4.2.1), the CR TEDE analysis considers the following sources of radiation that will cause exposure to control room personnel:

Contamination of the control room atmosphere by the filtered CR ventilation inflow through the CR air intake and by unfiltered inleakage of the radioactive material contained in the post-accident radioactive plume released from the facility, Contamination of the control room atmosphere by filtered CR ventilation inflow via the CR air intake and by unfiltered inleakage of airborne radioactive material from areas and structures adjacent to the control room envelope, Radiation shine from the external radioactive plume released from the facility (i.e., external airborne cloud shine dose),

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

[i.e., CR filter shine dose].

The external airborne cloud shine dose, containment shine dose, and the CR filter shine dose due to a LOCA are analyzed in Section 3.6.

Nuclear Common Revision 9

I CALCULATION CONTINUATION SHEET ISHEET 23 of 78 CALC. NO.: S-C-ZZ-MDC-1945

REFERENCE:

LCR S03-05 l

G. Patel/NUCORE, ORIGINATOR, DATE I REV:

02/24/04 0

l l

M Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/26/04 4.9.2 Consistent with RG 1.183 (Ref. 9.1, Section 4.2.2), the radioactive material releases and radiation levels used in the control room dose analysis are determined using the same source term, transport, and release assumptions used for determining the EAB and the LPZ TEDE values. These parameters do not result in non-conservative results for the control room.

4.9.3 Consistent with RG 1.183 (Ref. 9.1, Section 4.2.3), the models used to transport radioactive material into and through the control room, and the shielding models used to determine radiation dose rates from external sources, are structured to provide suitably conservative estimates of the exposure to control room personnel.

4.9.4 Consistent with RG 1.183 (Ref. 9.1, Section 4.2.4), credit for engineered safety features (ESF) that mitigate airborne radioactive material within the control room is assumed. Such features include control room pressurization, and intake and recirculation filtration. CR isolation is actuated by ESF signals and radiation monitors (RMs). Several aspects of CREACS operation can delay the CR isolation. A conservative delay of 1 minute is assumed for the CR isolation to be fully operational (Ref. 9.42, page 15).

4.9.5 Consistent with RG 1.183 (Ref. 9.1, Section 4.2.5), credit is not taken for the use of personal protective equipment (e.g., protective beta radiation resistant clothing, eye protection, or self-contained breathing apparatus [SCBA]) or prophylactic drugs (i.e., potassium iodide [KI] pills).

4.9.6 Consistent with RG 1.183 (Ref. 9 9.1, Section 4.2.6), the CR dose receptor for these analyses is the hypothetical maximum exposed individual who is present in the control room for 100% of the time during the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the event, 60% of the time between 1 and 4 days, and 40% of the time from 4 days to 30 days. For the duration of the event, the breathing rate of this individual should be assumed to be 3.5 x 1 0 4 cubic meters per second. These assumptions are incorporated as design inputs in Sections 5.5.12 and 5.5.14, respectively.

4.9.7 Consistent with RG 1.183 (Ref. 9.1, Section 4.2.7), the control room doses are calculated using the offsite dose analysis dose conversion factors identified in RG 1.183, Section 4.1. The DDE from photons is corrected for the difference between finite cloud geometry in the control room and the semi-infinite cloud assumption used in calculating the dose conversion factors. The RADTRAD Code (Ref. 9.2) used in this analysis uses the following expression to correct the semi-infinite cloud dose, DDE., to a finite cloud dose, DDEr,1,t,, where the control room is modeled as a hemisphere that has a volume, V, in cubic feet, equivalent to that of the control room:

DDEfinite = (DDEa) x v 0.3 3 8) / 1173 4.9.8 Consistent with RG 1.183 (Ref. 9.1, Section 4.4), for the LOCA the postulated CR doses should not exceed the 5 Rem TEDE criterion established in 10 CFR 50.67 (Ref. 9.36). This assumption is incorporated as a design input in Section 5.5.13:

CR Dose Acceptance Criterion:

5 Rem TEDE l Nuclear Common Revision 9

I CALCULATION CONTINUATION SHEET ISHEET 24 of 78 CALC. NO.: S-C-ZZ-MDC-1945

REFERENCE:

LCR S03-05 G. Patel/NUCORE, ORIGINATOR, DATE IREV:

02/24/04 0

l l

M Drucker/NUCORE, REVIEWERIVERIFIER, DATE 02/26/04 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 SNGS specific design inputs and assumptions used in the TID-14844 analyses were assessed for their validity to represent the as-built condition of the plant and evaluated for their compatibility to meet the AST and TEDE methodology. The analysis in this calculation ensures that analysis assumptions, design inputs, and methods are compatible with the requirements of the AST and the TEDE criteria.

5.1.2 Credit for Engineered Safety Features Credit is taken only for those accident mitigation features that are classified as safety-related, are required to be operable by technical specifications, are powered by emergency power sources, and are either automatically actuated or, in limited cases, have actuation requirements explicitly addressed in emergency operating procedures. The operation of 2 out of 5 containment fan cooler units, a delay of 1 minute in the CR pressurization, and operation of one train of CREACS are conservatively assumed to maximize the resulting doses. The CR intake radiation monitor capability to align with the less contaminated air intake is credited in the analysis.

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. The use of a 10% higher flow rate for the CREACS intake and a 10% lower recirculation flow rate, 1 minute delay in the CREACS initiation time, operation of one train of CREACS, and use of ground release X/Qs demonstrate the inherent conservatisms in the plant design and post-accident response. Most of the design input parameter values used in the analysis are those specified in the Technical Specifications (Ref. 9.6).

5.1.4 Meteorology Considerations Atmospheric dispersion factors (x/Qs) for the onsite release points such as the plant vent for containment and ESF leakage release paths are re-established using the NRC sponsored computer code ARCON96 (Ref. 9.5). The EAB and LPZ x/Qs are reconstituted using the SGS plant specific meteorology and appropriate regulatory guidance (Ref. 9.10). The site boundary X/Qs reconstituted in Reference 9.10 were accepted by the staff in the previous licensing proceedings.

I Nuclear Common Revision 9 I Nuclear Common Revision 9 I

I CALCULATION CONTINUATION SHEET SHEET 25 of 78 CALC. NO.: S-C-ZZ-MDC-1945

REFERENCE:

LCR S03-05 G. Patel/NUCORE, ORIGINATOR, DATE IREV:

02t24/04 0

M Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/26/04 5.2 Accident-Specific Design Inputs/Assumptions The design inputs/assumptions utilized in the EAB, LPZ, and CR habitability analyses are listed in the following sections. The design inputs are compatible with the requirements of the AST and TEDE dose criteria and the assumptions are consistent with those identified in Section 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.

I Nuclear Common Revision 9 Nuclear Common Revision 9

CALCULATION CONTINUATION SHEET SHEET 26 of 78 CALC. NO.: S-C-ZZ-MDC-1945

REFERENCE:

LCR S03-05 G. Patel/NUCORE, ORIGINATOR, DATE REV:

02/24/04 0

M DruckerlNUCORE, REVIEWER/VERIFIER, DATE 02/26/04 Design Input Parameter Value Assigned l

Reference 5.3 Containment Leakage Parameters 5.3.1 Source Term 5.3.1.1 Rated Thermal Power 3,459 MWt 9.6.1 3,632 MWt (= 105% of 3,459 MWt)

Used in the analysis 5.3.1.2 Isotopic Core Inventory (Ci/MWt) (Table 3)

Isotope Ci/MWt Isotope Ci/MWt Isotope Ci/MWt CO-58 2.553E+02 RU-103 3.598E+04 CS-136 1.042E+03 CO-60 1.953E+02 RU-105 2.340E+04 CS-137 1.915E+03 KR-85 3.056E+02 RU-106 8.175E+03 BA-139 4.976E+04 KR-85M 7.222E+03 RH-105 1.621E+04 BA-140 4.924E+04 KR-87 1.306E+04 SB-127 2.208E+03 LA-140 5.032E+04 KR-88 1.861E+04 SB-129 7.820E+03 LA-141 4,615E+04 RB-86 1.496E+01 TE-127 2.132E+03 LA-142 4.449E+04 SR-89 2.844E+04 TE-127M 2.823E+02 CE-141 4.476E+04 SR-90 1.535E+03 TE-129 7.341E+03 CE-143 4.352E+04 SR-91 3.656E+04 TE-129M 1.935E+03 CE-144 2.697E+04 SR-92 3.805E+04 TE-131M 3.707E+03 PR-143 4.273E+04 Y-90 1.647E+03 TE-132 3.690E+04 ND-147 1.911 E+04 Y-91 3.465E+04 1-131 2.750E+04 NP-239 5.120E+05 Y-92 3.819E+04 I-132 3.889E+04 PU-238 2.902E+01 Y-93 4.320E+04 1-133 5.556E+04 PU-239 6.545E+00 ZR-95 4.377E+04 1-134 6.11 lE+04 PU-240 8.254E+00 ZR-97 4.562E+04 1-135 5.278E+04 PU-241 1.390E+03 NB-95 4.138E+04 XE-133 5.556E+04 AM-241 9.181E-01 MO-99 4.830E+04 XE-135 1.389E+04 CM-242 3.514E+02 TC-99M 4.169E+04 CS-134 3.425E+03 CM-244 2.056E+01 5.3.1.3 Radionuclide Composition Group Elements Noble Gases Xe, Kr 9.1, Table 5 Halogens I, Br Alkali Metals Cs, Rb Tellurium Group Te, Sb, Se, Ba, Sr Noble Metals Ru, Rh, Pd, Mo, Tc, Co Lanthanides La, Zr, Nd, Eu, Nb, Pm, Pr, Sm, Y, Cm, Am Cerium Ce, Pu, Np 5.3.1.4 Timing of [PWR] Release Phases (Ref. 9.1, Table 4)

Phase Onset Duration Gap Release 30 sec 0.5 hr Early In-Vessel Release 0.5 hr 1.3 hr Nuclear Common Revision 9

CALCULATION CONTINUATION SHEET SHEET 27 of 78 CALC. NO.: S-C-ZZ-MDC-1945

REFERENCE:

LCR S03-05 G. Patel/NUCORE, ORIGINATOR, DATE REV:

02/24/04 0

M Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/26/04 Design Input Parameter l

Value Assigned Reference 5.3.1.5 Fraction of [PWR] Core Inventory Released Into Containment (Ref 9.1, Table 2)

Group Gap Release Phase Early In-Vessel Release Phase Noble Gases 0.05 0.95 Halogens 0.05 0.35 Alkali Metals 0.05 0.25 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 Iodine Chemical Form Iodine Chemical Form Aerosol (CsI) 95.0%

9.1, Section 3.5 Elemental 4.85%

Organic 0.15%

5.3.2 Activity Transport in Primary Containment 5.3.2.1 Containment Net Free Volume 2,620,000 ft 9.6.10 5.3.2.2 Sprayed Volume 75% of containment volume 9.34, Section 3.4.1.4 and 1.965E+06 ft3 Section 3.2.1 5.3.2.3 Unsprayed Volume 25% of Containment volume 6.55E+05 ft3 5.3.2.4 Containment Leak Rate 0.1 v%/day 0-24 Hrs 9.6.4 & 9.7, Section 4.13 0.05 v%/day 24-720 Hrs 9.1, App A, Section 3.7 5.3.2.5 Containment Fan Cooler Unit 39,000 cfm/fan 9.14, Section 2.2.d (CFCU) Flow Rate 5.3.2.6 Flow Rate From Sprayed To 21,833 cfm (2 of 5 CFCUs operational)

Section 3.2.1 Unsprayed Volume 5.3.2.7 Flow Rate From Unsprayed To 21,833 cfm (2 of 5 CFCUs operational)

Section 3.2.1 Sprayed Volume 5.3.2.8 Spray Initiation Time 85 sec 9.12, page 95 90 sec used in the analysis 5.3.2.9 Spray Recirculation Phase 60 min used in iodine transport analysis 9.8, Section 4.1 Initiation Time 48 min used in containment leakage 9.11 analysis 5.3.2.10 Maximum Iodine Activity 1.8 hr 9.1, Section 3.3 Occurs in Containment (Assumed coincident with the end of the Early In-Vessel release phase)

Nuclear Common Revision 9

CALCULATION CONTINUATION SHEET SHEET 28 of 78 CALC. NO.: S-C-ZZ-MDC-1945

REFERENCE:

LCR S03-05 G. Patel/NUCORE, ORIGINATOR, DATE REV:

02/24/04 0

M Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/26/04 Design Input Parameter Value Assigned Reference 5.3.2.11 Containment Spray Injection 5,200 gpm (2 trains at 2,600 gpm) used in 9.8, Table 4-2 Phase Flow Rates iodine transport analysis 2,600 gpm (1 train at 2,600 gpm) used in containment leakage analysis 5.3.2.12 Spray Removal Coefficients In Injection Phase Elemental (XEs2oo) 58 h i' (= 2 x 29 hi') used in iodine 9.8, Table 4-2 transport analysis Elemental (E2600) 20 hl' used in containment leakage 9.9, page 6.5.2-11 analysis Particulate (XP5200) (Aerosol) 8.88 hf ' (= 2 x 4.44 hlif) used in iodine Section 6.2 transport analysis Particulate (XP2600) (Aerosol) 4.44 hId used in containment leakage analysis 5.3.2.13 Containment Spray Flow Rate 1,900 gpm used in iodine transport 9.34, Section 3.4.4 During Recirculation Phase analysis and containment leakage analysis 5.3.2.14 Spray Removal Coefficients In Recirculation Phase Elemental (E1900) 21.19 hi' used in iodine transport analysis Section 6.3 14.62 hl' used in containment leakage analysis Particulate (Xplooo) (Aerosol) 3.24 hi' used in both iodine transport and Section 6.3 containment leakage analyses 5.3.2.15 Decontamination Factor (DF) For Spray Cutoff Time Elemental 200 used in iodine transport analysis 9.9, Section 4.c.(4).d 100 used in containment leakage analysis Particulate (Aerosol) 50 used in both iodine transport and containment leakage analyses 5.3.2.16 Spray Interruption During 5 minutes Assumed Transition From Injection To 10 minutes used in the analysis Recirculation Phase 5.3.2.17 Post-LOCA Spray Removal Coefficients (hri') During Gap and Early-In-Vessel Release Phases for Iodine Transport Analysis (Sections 6.2 through 6.4)

Spray Removal Coefficients (hi'1)

Time (hour)

Aerosol Elemental 0

0 0

0.025 (90 seconds) 8.88 58 1.000 (60 minutes) 3.24 21.19 2.050 3.24 0

3.007 0.324 0

4.0 0

0 Nuclear Common Revision 9

lCALCULATION CONTINUATION SHEET l SHEET 29 of 78 CALC. NO.: S-C-ZZ-MDC-1945

REFERENCE:

LCR S03-05 G. Patel/NUCORE, ORIGINATOR, DATE REV:

02/24/04 0

M Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/26/04 Design Input Parameter Value Assigned l

Reference 5.3.2.18 Post-LOCA Spray Removal Coefficients (hf ') During Gap and Early-In-Vessel Release Phases for Containment Leakage Analysis (Sections 6.2 through 6.4)

Time (hour)

Aerosol Elemental 0

0 0

0.025 (90 seconds) 4.44 20 0.800 (48 minutes) 0 0

0.967 (58 minutes) 3.24 14.62 2.115 3.24 0

3.007 0.324 0

4.0 0

0 5.4 ESF Leakage Parameters 5.4.1 Sump Water Volume 43,930 ft3 @ > 48 minutes 9.15 5.4.2 ESF Leakage Rate 3790 cc/hr (= 1.0 galfhr) 9.16, page 10 1 gpm (2 gpm used in the analysis) 9.1, App A, Section 5.2 5.4.3 ESF Leakage Initiation Time 20 minutes (earliest possible time with 9.8, page 4-1 ECCS operating at maximum capacity) 5.4.4 Sump Water pH

> 7.0 9.37, Section 5.0 5.4.5 ESF Leakage Iodine Flashing 10%

9.1, App A, Section 5.5 Factor 5.4.6 Chemical Form of Iodine In ESF Leakage Elemental 97%

9.1, App A, Section 5.6 Organic 3%

5.4.7 Fraction of Core Elemental 0.05824 Table 11 Iodine In Sump Water 5.4.8 Auxiliary Building Ventilation 0.0% (assumed for elemental, organic 9.34, Section 4.0 System (ABVS) Charcoal Filter and particulate iodine)

Efficiencies 5.4.9 Containment Temperature 271uF (design) 9.8, Table 4-2 5.5 Control Room (CR) Parameters 5.5.1 CRVolume 81,420ft 9.18, page 33 5.5.2 CR Normal Flow Rate 1,320 cfm (for two air intakes)

Section 6.5 5.5.3 CREACS Makeup Rate 2,200 cfm 9.6.3 5.5.4 CREACS Recirc Flow Rate 8,000 cfm +/- 10%

9.6.2 5,000 cfm (used in analysis)

Section 6.5 5.5.5 CREACS Charcoal Filter 95% for elemental iodine Section 6.6 Efficiencies 95% for organic iodide 5.5.6 CREACS HEPA Filter 99%

Section 6.6 Efficiency Nuclear Common Revision 9l

CALCULATION CONTINUATION SHEET SHEET 30 of 78 CALC NO.: S-C-ZZ-MDC-1945

REFERENCE:

LCR S03-05 G. Patel/NUCORE, ORIGINATOR, DATE REV:

02/24/04 0

M Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/26/04 Design Input Parameter Value Assigned Reference 5.5.7 CR Unfiltered Inleakage 150 cfhn (nominal value measured is less 9.33, Table 1 Determined By Tracer Gas Testing than 100 cfin) 9.34, Section 3.2.5 5.5.8 Unit 1 CR x/Qs For Containment & ESF Leakage Release Via Unit 1 Plant Vent Time X/Q (see/mr) 0-2 1.78E-03 9.5, page 34 2-8 1.3 1E-03 8-24 5.22E-04 24-96 3.77E-04 96-720 3.17E-04 5.5.9 Unit 2 CR X/Qs For Containment & ESF Leakage Release Via Unit 1 Plant Vent Time X/Q (sec/mr) 0-2 8.84E-04 9.5, page 34 2-8 6.60E-04 8-24 2.64E-04 24-96 1.93E-04 96-720 1.62E-04 5.5.10 CR Wall Thickness 2.0 feet 9.23.a 5.5.11 CR Roof Thickness 2.0 feet 9.23.c 5.5.12 CR Breathing Rate 3.5E-04 m3/sec 9.1, Section 4.2.6 5.5.13 CR Allowable Dose Limit 5 rem TEDE for the event duration 9.36 5.5.14 CR Occupancy Factors Time (Hr)

__9.1, Section 4.2.6 0-24 100 24-96 60 96-720 40 5.6 Site Boundary Release Model Parameters 5.6.1 EAB Atmospheric Dispersion 1.30E-04 sec/Mr3 9.10, Table 5 Factor (X/Q) 5.6.2 EAB Breathing Rate (m'/sec) 3.5E-04 9.1, Section 4.1.3 5.6.3 LPZ Atmospheric Dispersion Factors (x/Qs)

Time (Hr)

XIQ (scc/m')

9.10, Table 5 0-2 1.86E-05 2-8 7.76E-06 8-24 5.01E-06 24-96 1.94E-06 96-720 4.96E-07 Nuclear Common Revision 9

I CALCULATION CONTINUATION SHEET I SHEET 31 of 78 CALC.NO.: S-C-ZZ-MDC-1945

REFERENCE:

LCR S03-05 G. Patel/NUCORE, ORIGINATOR, DATE REV:

02/24/04 0

M Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02126/04 Design Input Parameter Value Assigned Reference 5.6.4 Offsite [Low Population Zone (LPZ)] Breathing Rate (m3/sec)

Time (Hr)

(m/sec) 9.1, Section 4.1.3 0-8 3.5E-04 8-24 1.8E-04 24-720 2.3E-04 5.6.5 EAB allowable dose limit 25 rem TEDE for any 2-hour period 9.1, Table 6 9.36 5.6.6 LPZ allowable dose limit 25 rem TEDE for the event duration 9.1, Table 6 9.36 5.7 Containment Vacuum Relief Line Release Parameters 5.7.1 Relief Valve Closure Time 5.0 sec 9.6.9 5.7.2 Volume of Containment Air 1,014 ftW 9.24, page 9 Release Via Vacuum Relief Line 5.7.3 Reactor Coolant System Volume 12,446 ft3 @ a nominal Tayg of 573uF J 9.6.11 5.7.4 Maximum Coolant (1% Failed Fuel) Reactor Activity Values (Ref. 9.3, Table 4)

Isotope ptCi/g Isotope PCi/g Isotope t'Ci/g KR-83M 4.0E-01 XE-133M 1.7E+01 I-131 2.8E+00 KR-85M 1.7E+00 XE-133 2.6E+02 1-132 2.8E+00 KR-85 8.2E+00 XE-135M 4.9E-01 1-133 4.2E+00 KR-87 1.0E+00 XE-135 8.5E+00 1-134 5.7E-01 KR-88 3.OE+00 XE-137 1.8E-01 1-135 2.3E+00 XE-131M 2.1E+00 XE-138 6.1E-01 5.8 RWST Back-leakage Parameters 5.8.1 RWST Back-leak Rate 1 00 cc/hr 9.13, page 3 5.8.2 RWST Minimum Air Volume 35,806 ftE 9.13, page 3 INuclear Common Revision 9 I Nuclear Common Revision 9 I

CALCULATION CONTINUATION SHEET SHEET 32 of 78 CALC. NO.: S-C-ZZ-MDC-1945

REFERENCE:

LCR S03-05 G. Patel/NUCORE, ORIGINATOR, DATE l REV:

02/24/04 0

M DruckerfNUCORE, REVIEWERNVERIFIER, DATE 02126/04

6.0 CALCULATIONS

6.1 Salem 1 & 2 Plant Specific Nuclide Inventory File (NIF) For RADTRAD3.02 Input The RADTRAD nuclide inventory file Pwrdef.NIF establishes the power dependent radionuclide activity in Ci/MWt for a typical PWR reactor core inventory. The aerosol nuclide inventory in this PWR default file is combined with the Salem plant-specific iodine and noble gas nuclide inventory to generate a Salem plant specific NIF SalemDEF.NIF. The Salem plant-specific iodine and noble gas nuclide inventory is developed using methodology discussed in Section 3.1 and shown in Tables 1 through 3.

This methodology considers the Salem plant-specific core thermal power level and those iodine and noble gas radionuclides that are significant dose contributors. The newly developed NIF SalemDEF.NIF is used with the RADTRAD PWR design basis accident Release Fraction & Timing (RFT) file Pwrdba.rft and Dose Conversion File Fgrl 1&12.inp in the AST analysis.

6.2 Containment Spray Removal Coefficients - Injection Phase:

Elemental Iodine Removal Coefficient for Containment Leakage Analysis The containment leakage analysis is modeled with a single train of containment spray operating at 2600 gpm during the injection phase. The elemental iodine spray removal coefficient is XE = 29 hf' at 2600 gpm (Ref 9.8, Table 4-2). However, Standard Review Plan (SRP) 6.5.2 limits XE = 20 hf ' for use in the AST containment leakage analysis (Ref. 9.9, page 6.5.2-11).

Particulate (Aerosol) Removal Coefficient for the Containment Leakage Analysis The Particulate (Aerosol) spray removal coefficient Xp is calculated using the SRP 6.5.2 methodology and plant-specific information. The first order removal coefficient Xp, for aerosol may be estimated by:

Xp = 3hFE (Ref. 9.9, page 6.5.2-11) 2VD Where h = fall height of the spray drops F = volumetric flow rate of spray pump per hr (E/D) = ratio of a dimensionless efficiency E to the average spray drop diameter D = 10 per meter V = containment building net free volume These parameters have the following plant-specific values:

h = 122.0 ft (Ref. 9.8, Table 4-2)

F = 2,600 gal/min (Ref. 9.8, Table 4-2)

= 2,600 gal/min x (1 / 7.481 ft3/gal) x 60 min/hr = 20,853 ft3/hr (E/D) = 10 m x (1 /3.28 ft/m) = 3.049 ft-1 I Nuclear Common Revision 9 1 Nu l a Co m nR v s o

CALCULATION CONTINUATION SHEET SHEET 33 of 78 CALC. NO.: S-C-ZZ-MDC-1945

REFERENCE:

LCR S03-05 G. Patel/NUCORE, ORIGINATOR, DATE REV:

02/24/04 0

M Drucker/NUCORE, REVIEWERJVERIFIER, DATE 02/26/04 V = 2.62E+06 ft3 (Ref. 9.6.10)

Substituting these values into the above equation for Xp:

%2600 = 3hFE = 3 x 122.0 ft x 20,853 ft3/hr x 3.049 f'l = 4.44 hi' 2VD 2 x 2.62E+06 ft3 Elemental and Particulate Iodine Removal Coefficients for the Iodine Transport To Sump Analysis The iodine transport to sump analysis is modeled with two trains of containment spray each operating at 2600 gpm during the injection phase. The elemental and particulate iodine spray removal coefficients for a single train of containment spray operating at 2600 gpm are:

XE2600 = 29 hi' (Ref. 9.8, Table 4-2)

?P2600 = 4.44 hl' (preceding Section 6.2 calculation)

With two trains of containment spray each operating at 2600 gpm, the elemental and particulate iodine removal rates for a single train are effectively doubled. The resultant elemental and particulate iodine spray removal coefficients for the iodine transport to sump analysis are:

XEs2oo =2 x XE2600 =2 x 29 hli = 58 hF1 XP5200 = 2 x XP2600 = 2 x 4.44 hril = 8.88 hr4l 6.3 Spray Removal Coefficients During Recirculation Phase Per Section 6.2, the spray removal coefficient is a function of the spray drop fall height, spray flow rate, average spray drop diameter and containment building net free volume. The spray drop fall height and containment building volume are not impacted by a change from the spray injection phase to recirculation phase. It is assumed that the spray mass-mean diameter is not changed significantly by the spray injection phase to recirculation phase flow rate reduction from 2,600 gpm to 1,900 gpm.

Containment spray injection phase coverage is calculated at the containment pressure of 47 psig and temperature of 271°F (Ref. 9.8, Table 4-2). The spray trajectories for the given size of drop are expected to change in the environments of different densities as shown in Figure 2-8 of Reference 9.8. It is assumed that the effective spray coverage of 75% is maintained during recirculation phase because the effect of reduced spray flow rate of 1900 gpm on the spray coverage will be compensated by the increased trajectories of spray drops resulting from the reduced back pressure at the spray nozzles and the lower density of containment air during the recirculation phase. The elemental and particulate removal coefficients are assumed to vary directly in proportion to the spray flow rate. The spray removal coefficients in recirculation mode are calculated as follows:

Nuclear Common Revision 9 I Nuclear Common Revision 9 l

REVIEWER/VERIFIER, DATE 0226/04 I

I Elemental Iodine and Particulate (Aerosol) Removal Coefficients for Containment Leakage Analysis The containment leakage analysis is modeled with a single train of containment spray operating at 1900 gpm during the recirculation phase (Ref. 9.34, Section 4.0).

The elemental and particulate iodine spray removal coefficients for a single train of containment spray operating at 2600 gpm during the injection phase are:

XE2600 = 20 hl' (Section 6.2)

XP2600 = 4.44 hI' (Section 6.2)

With one train of containment spray operating at a reduced flow rate of 1900 gpm during the recirculation phase, the elemental iodine and particulate (aerosol) removal rates are scaled down to:

XE1900 = 1,900 gpm/2,600 gpm x 20.0 hi' = 14.62 hi' XP19oo = 1,900 gpm/2,600 gpm x 4.44 hl' = 3.24 hf '

Elemental and Particulate Iodine Removal Coefficients for the Iodine Transport To Sump Analysis The iodine transport to sump analysis is modeled with one train of containment spray operating at 1900 gpm during the recirculation phase (Ref. 9.34, Section 4.0).

The elemental and particulate iodine spray removal coefficients for a single train of containment spray operating at 2600 gpm during the injection phase are:

%E2600 = 29 hi' (Ref. 9.8, Table 4-2)

XP2600 = 4.44 hi1 (Section 6.2)

With one train of containment spray operating at a reduced flow rate of 1900 gpm during the recirculation phase, the elemental and particulate iodine removal rates are scaled down to:

XE1900 = 1,900 gpm/2,600 gpm x 29.0 hi'1 = 21.19 hi'

%P1soo = 1,900 gpm/2,600 gpm x 4.44 hi' = 3.24 hi' 6.4 Spray Removal Cutoff Times During Gap & Early-In-Vessel Release Phases 6.4.1 Injection Phase Spray Removal Cutoff Times Containment Leakage Analysis One spray pump operates @ 2,600 gpm from 90 sec to 48 minutes with XE2600 = 20 hi' (Section 6.2) and XP2600 = 4.44 hi ' (Section 6.2).

I Nuclear Common Revision 9 l Nuclear Common Revision 9

I CALCULATION CONTINUATION SHEET ISHEET 35 of 78 CALC. NO.: S-C-ZZ-MDC-1945

REFERENCE:

LCR S03-05 G. Patel/NUCORE, ORIGINATOR, DATE IREV:

02/24/04 0

l M Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/26/04 Iodine Transport To Sump Analysis Two spray pumps operate @

2,600 gpm each from 90 sec to 60 minutes with XES200 = 58 hf' (Section 6.2) and XP5200 = 8.88 hl' (Section 6.2).

6.4.2 Recirculation Phase Spray Removal Cutoff Times 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 1.8 hr after the onset of a LOCA (Ref. 9.1, Table 4).

Elemental Iodine Spray Removal Cutoff Time For Containment Leakage Analysis One spray pump operates ( 1,900 gpm from 48 minutes to 4.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> in the following manner:

DF for spray cutoff time for elemental iodine = 100 (conservatively assumed)

Containment leakage analysis elemental iodine spray removal coefficient (XE,900) = 14.62 hl' (per Section 6.3) 1/DF = 1/100 = 0.01 = e'(EI9 oo) t = el 4.62) t In (0.01)= - (14.62)t

- 4.605 = - (14.62)t Therefore spray cutoff time for elemental iodine t = -4.605/-14.62 = 0.315 hr Spray termination time for elemental iodine = 1.8 hr + 0.315 hr = 2.115 hr Elemental Iodine Spray Removal Cutoff Time For Iodine Transport To Sump Analysis One spray pump operates @

1,900 gpm from 60 minutes to 4.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> in the following manner:

DF for spray cutoff time for elemental iodine = 200 (Ref. 9.9, page 6.5.2-12)

Iodine transport to sump analysis elemental iodine spray removal coefficient (XEI900) = 21.19 hr' (Section 6.3) 1/DF = 1/200 = 0.005 = e7(X Ei9o)t = e-(21.19)t ln (0.005) = - (21.19)t

- 5.298 = - (21.19)t Therefore spray cutoff time for elemental iodine t = - 5.298/-21.19 = 0.250 hr Spray termination time for elemental iodine = 1.8 hr + 0.250 hr = 2.050 hr Particulate (Aerosol) Spray Removal Cutoff Time for both Containment Leakage and Iodine Transport to Sump Analyses One spray pump operates ( 1,900 gpm from either 48 or 60 minutes to 4.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> in the following manner:

I Nuclear Common Revision 9 1 Nu l a Co m nR v s o

CALCULATION CONTINUATION SHEET SHEET 36 of 78 CALC.NO.: S-C-ZZ-MDC-1945

REFERENCE:

LCR S03-05 G. Patel/NUCORE, ORIGINATOR, DATE REV:

02/24/04 0

M Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/26/04 The particulate removal rate should be reduced by a factor of 10 when a DF of 50 is reached (Ref. 9.1, Appendix A, Section 3.3, and Ref. 9.9, page 6.5.2-12).

Spray particulate removal coefficient (XPl900) = 3.24 hrf (Section 6.3) 1/DF = 1/50 = 0.02 = epigoo)t = e (3 4 ) t In (0.02) = - (3.24)t

-3.912 = - (3.24)t Therefore spray particulate removal rate reduction time t = -3.912/-3.24 = 1.207 hr Total spray particulate removal rate reduction time = 1.8 hr + 1.207 hr = 3.007 hr Therefore, the particulate removal coefficient of 3.24 hr' is used from either 48 or 60 minutes until 3.007 hrs, and the reduced particulate removal coefficient of 0.324 hr 4-is used from 3.007 hrs to 4.0 hrs, when the spray operation is assumed to terminate.

6.5 Control Room Flow Rates 6.5.1 Normal Flow Rate Notes "S" to the Reference 9.25 ventilation drawings provide the outside air flow rates to Zone 1 from the Unit 1 and Unit 2 air intakes. Zone 1 is the combined control room envelop. Zone 1 receives only a fraction of the 2,200 cfm of outside air introduced into the building. The fraction is equivalent to the ratio of the Zone 1 (control room pressure boundary) supply air flow rate [8,000 cfm] to the total control area air conditioning system (CAACS) normal airflow rate [32,600 cfm = 2,200 cfm outside air +

30,400 cfm recirculation air)]. Notes "S" provide the following calculation for the amount of outside air to Zone 1:

= (8,000 cfm / 32,600 cfm) x (2,200 cfm) = 540 cfm Per Notes "S", use 600 cfm for Zone I During Normal Plant Operation Total Amount of Outside Air Flow Rate From Both Intakes = 2 x 600 cfm = 1,200 cfm Maximum Amount of Outside Air Flow Rate = 1.1 x 1,200 cfm = 1,320 cfm (including 10 percent uncertainty) 6.5.2 CREACS Recirculation Flow Rate With CR Monitors Preferentially Selecting Less Contaminated Air Intake CREACS ventilation flow rate = 8,000 cfm + 10% (Ref. 9.6.2)

Minimum CREACS flow rate = 8,000 cfrn - (0.10 x 8,000 cfm) = 8,000 cfm - 800 cfm = 7,200 cfm Net CREACS recirculation flow rate = Minimum CREACS flow rate - CREACS makeup flow rate

= 7,200 cfm - 2,200 cfm (Ref. 9.6.3) = 5,000 cfm 6.6 CREACS Charcoal/HEPA Filter Efficiencies 6.6.1 Charcoal Filter In-place penetration testing acceptance criteria for the safety related Charcoal filters are as follows:

CREACS Charcoal Filter - in-laboratory testing methyl iodide penetration < 2.5% (Ref. 9.6.5)

I Nuclear Common Revision 9 1 Nuclear Common Revision 9

CALCULATION CONTINUATION SHEET SHEET 37 of 78 CALC. NO.: S-C-ZZ-MDC-1945

REFERENCE:

LCR S03-05 G. Patel/NUCORE, ORIGINATOR, DATE REV:

02/24/04 0

M Drucker/NUCORE, REVIEWERIVERIFIER, DATE 02/26/04 Generic Letter 99-02 (Ref. 9.26) requires a safety factor of at least 2 to be used to determine the filter efficiencies to be credited in the design basis accident.

Testing methyl iodide penetration (%) = (100% - #)/safety factor = (100% - 11)/2 Where -q = charcoal filter efficiency to be credited in the analysis CREACS Charcoal Filter 2.5% = (100% -)/2 5%= (100%- I)

' = 100% - 5% = 95%

6.6.2 HEPA Filter HEPA filter efficiency = 99% (Ref. 9.6.6). HEPA filter efficiency of 99% is used in the analysis Safety Grade Filter Efficiency Credited (%)

Filter Aerosol Elemental Organic CREACS 99 95 95 6.7 RCS Activity Released During Containment Vacuum Relief The containment vacuum relief line release evaluation assumes that the entire RCS water inventory including the iodine activity at the Technical Specification concentration of 1 piCi/g Dose Equivalent I-131 is instantaneously released and homogeneously mixed in the containment atmosphere. The RCS mass is calculated as follows:

RCS water volume = 12,446 R3 ( a nominal Tavj of 573 0F (Ref. 9.6.11)

RCS water specific volume @ 573 0F = 0.02253 ft Abm (Ref. 9.35, page 182)

RCS water mass = (12,446 f3 / 0.02253 ft3/lbm) x 453.6 g/lbm = 2.5 1E+08 g Volume of containment air released = 1014 ft3 (Ref. 9.24, page 9)

Duration of release = 5.0 sec (Ref. 9.6.9)

Vacuum relief release rate = (1014 ft3 / 5.0 sec) x 60 sec/min = 12,170 cfm The RCS isotopic activities released into the containment are calculated in Tables 4 through 7 and the RADTRAD NIF is developed in Table 8, using the methodology described in Section 3.4. The RADTRAD NIF for the containment vacuum relief line release is modified based on Ci/MWt information developed in Table 8. A new RADTRAD NIF Spurge def.txt is developed. The RADTRAD dose conversion file (DCF) Fgrl 0&11.inp is modified to include short-lived noble gas isotopes. The newly developed DCF Spurge fgl 1&12.txt is used to calculate the CR and site boundary doses.

I Nuclear Common Revision 9 1 Nu l a Co m nR v s o

I CALCULATION CONTINUATION SHEET ISHEET 38 of 78 CALC. NO.: S-C-ZZ-MDC-1945

REFERENCE:

LCR S03-05 G. Patel/NUCORE, ORIGINATOR, DATE IREV:

02/24/04 0

l l

M Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02126/04 6.8 Salem I Plant Vent Release - Center of Salem 1 CR X/Qs for CR External Cloud Dose The location of the Unit 1 Plant Vent (PV) with respect to the center of the Unit I CR is shown in Figure 5 (Ref. 9.28). The PV location with respect to the CR center is such that the south wind will predominantly carry effluent from the PV. The containment building cross-sectional area perpendicular to the south wind is considered for the wake diffusion.

Containment cross-sectional area perpendicular to wind direction = 2,429.54 m2 (Ref. 9.5, Section 7.1)

The PVI assumed to be located at the center and top of the containment bldg at EL 291'-6" (Ref. 9.28.d)

Center of Unit 1 CR is located @

Column 13.2 and Row DD (Ref. 9.22.a)

North-south distance between centerlines of containment (PV) and center of CR

= 97-6" + 13'-6" + 1 1'-6" + 10'-6" = 133'-0" (Ref. 9.22.a)

East-west distance between centerline of PV and CR center = Distance between Rows FF & DD

= 14'-0" + 20'-7" = 34'-7" = 34.583' (Ref. 9.22.a)

Straight line distance between PV and CR center = [(133)2 + (34.583)2]1/2 = 137.42 ft = 41.90 m.

Elevation of CR roof = 140'-0" (Ref. 9.23.c)

Grade elevation = 99'-6" (Ref. 9.28.d)

Height of CR roof =140'-0" - 99'-6" = 40'-6" = 12.35 m Elevation of Plant Vent = 291 '-6" (Ref. 9.28.d)

Height of Release Point = 291 '-6" - 99'-6" = 192' = 58.54 m PV direction with respect to CR center Tan 0 = 34.583/133 = 0.26, Therefore 0 = Tan-i 0.26 = 14.58° Orientation of PV release with respect to CR center, considering south wind 1800 and true north wind 3600 (Ref. 9.43, page 16), and that plant north is 50-30'-01" east of true north [with 60 minutes per degree] (Ref. 9.46)

Orientation= 1800+ 14.58°- 5.500= 189.08° 6.9 Iodine/Aerosol Deposition on CR CharcoaJIHEPA Filter - Containment Leakage:

6.9.1 Iodine Activity Deposited On CR Charcoal Filter:

As shown in Figure 3, the CR intake and recirculation charcoal filter elemental & iodine removal efficiency is 95% with the intake and recirculation flow rates of 2,200 cfm and 5,000 cfm, respectively.

Charcoal Filter 2090 Iin Suppose Air Intake Iodine Activity (atoms) = Iin and Iodine Activity Deposited on Recirc Filter = Ide Activity deposited on the intake charcoal filter = Iin x 0.95 x 2,200 cfm = 2090 Iin Filtered inflow activity introduced into the CR = Iin x (1 - 0.95) x 2,200 cfm = 110 Iin I Nuclear Common Revision 9 1 I Nuclear Common Revision 9

I l CALCULATION CONTINUATION SHEET l SHEET 39 of 78 CALC. NO.: S-C-ZZ-MDC-1945

REFERENCE:

LCR S03-05 G. Patel/NUCORE, ORIGINATOR, DATE REV:

02/24/04 0

M Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/26/04 l 110 Ii, + 150 Iin (unfiltered inleakage) =260 Iin OPr Recirc Charcoal Filter 10 Tj = 95%

13 Iin Id, 247 Ii, Activity deposited on recirc charcoal filter (conservatively assuming only one pass through the recirc charcoal filter) = Id, = 0.95 x (260 Iin cfm) = 247 Iin; Rearranging: Iin = Ide /247 Therefore, the activity deposited on the intake charcoal filter = 2090 Iin= (2090/247) Ide Total elemental & organic iodine activity deposited on intake + recirc charcoal filter

= (2090/247 Ide) + Id: `e 9.5 Id, Id, = 1.2295E+15 Atoms (Elemental Iodine) + 6.1425E+14 Atoms (Organic Iodine)

(S15OCS75CLO01.o0, CR Recirculating Filter Nuclide Inventory @ 720 hrs)

= 1.84375E+15 Atoms Total iodine activity deposited on the CR intake + recirc charcoal filter due to containment leakage

= 9.5 x 1.84375E+15 Atoms = 1.7516+16 Atoms, which is used in Table 14 to obtain the iodine isotopic activities.

6.9.2 Aerosol Mass Deposited On CR HEPA Filter:

As shown in Figure 3, the CR intake and recirculation HEPA filter aerosol removal efficiency is 99% for the intake and recirculation flow rates of 2,200 cfm and 5,000 cfm respectively.

_ l Intake HEPA Filter 1

2200 Ai,

-q = 99%

22 Ain 2178 Ain Suppose aerosol mass in intake air = Ain and aerosol mass deposited on recirc filter = Ad, Aerosol mass deposited on HEPA filter = Ain x 0.99 x 2,200 cfm = 2178 Ain Filtered inflow aerosol mass introduced into the CR = Ain x (1 - 0.99) x 2,200 cfm = 22 Ain INuclear Common Revision 9 Nuclear Common Revision 9 I

I CALCULATION CONTINUATION SHEET l SHEET 40 of 78 CALC. NO.: S-C-ZZ-MDC-1945

REFERENCE:

LCR S03-05 G. Patel/NUCORE, ORIGINATOR, DATE IREV:

02/24/04 0

l l

M Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/26/04 l22 Ain + 150 Ain (unfiltered inleakage) =172 Ain v>

Recirc HEPA Filter 1.72 Ain rl = 99%

Ad, 170.28 Ai, Aerosol mass deposited on recirc HEPA filter (conservatively assuming only one pass through the recirc HEPA filter) = Ade = 0.99 x (172 Aincfm) = 170.28 Ain; Rearranging: Ain= Ade / 170.28 Therefore, the aerosol mass deposited on intake HEPA filter = 2178 Ain= (2178/170.28) Ad, Total aerosol mass deposited on intake + recirc HEPA filter

= (2178/170.28 Ade) + Ad, $w 13.80 Ad:

Ad, = 4.4653E-08 kg (S150CS75CLO1.o0, CR Recirculating Filter Nuclide Inventory @720 hrs)

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

= 13.80 x 4.4653E-08 kg = 6.162E-07 kg, which is used in Table 13 to obtain the aerosol isotopic activities.

6.10 Iodine/Aerosol Deposition on CR Charcoal/HEPA Filter - ESF Leakage:

6.10.1 Iodine Activity Deposited On CR Charcoal Filter:

As discussed in Section 6.9.1 above:

Total elemental & organic iodine activity (atoms) deposited on intake + recirc charcoal filter 9.5 Id, Id&

= 1.0019E+17 Atoms (Elemental Iodine) + 3.0986E+15 Atoms (Organic Iodine) (S75ESFIGPM.oO, CR Recirculating Filter Nuclide Inventory @

720 hrs)

= 1.0329E+17 Atoms.

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

= 9.5 x 1.0329E+17 Atoms = 9.813E+17 Atoms, which is used in Table 16 to obtain the iodine isotopic activities.

6.10.2 Aerosol Mass Deposited On CR HEPA Filter:

Since post-LOCA ESF leakage consists of only a non-aerosol iodine release (97% of elemental iodine +

3% of organic iodine) (Ref. 9.1, Section 5.6), there is no aerosol deposited on the CR intake + recirc HEPA filter (S 150ESF1 GPM.oO, CR Compartment Nuclide Inventory @

720 hrs)

I Nuclear Common Revision 9 1 Nuclear Common Revision 9 I

I CALCULATION CONTINUATION SHEET l SHEET 41 of 78 CALC. NO.: S-C-ZZ-MDC-1945

REFERENCE:

LCR S03-05 G. Patel/NUCORE, ORIGINATOR, DATE IREV:

02/24/04 0

l l

M Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/26/04 6.11 CR Direct Dose From Control Room Filter Shine:

CR Filter Shine Average Dose Rate during the LOCA event = 7.564E-01 mrem/hr (MicroShield Run SCREAS.MS5)

CR Operator Exposure Time

= 1 x (24 hr) + 0.60 (96 hr -24 hr) + 0.40 (720 hr -96 hr)

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

= 7.564E-01 mrem/hr x 1/1000 rem/mrem x 316.8 hr = 2.40E-01 rem Adjusted CR Filter Shine Dose for Increased Aerosol inventory = 1.1 x 2.40E-01 rem = 2.64E-01 rem, which is added to other post-LOCA dose contributions in Section 7.0 6.12 ESF Leakage Flashing Factor for Sump Temperature > 212 0F The flashing factor (FF) of the ESF leakage for sump water temperature greater than 2120F is determined using a constant enthalpy process per RG 1.183, Appendix A, Section 5.4 as follows:

FF = hfl

-_h2 hrg Where hn is the enthalpy of liquid at system design temperature and pressure; hf is the enthalpy of liquid at saturation conditions (14.7 psia, 212°F); and hfg is the heat of vaporization at 212°F.

Assuming sump water is in equilibrium with the containment at the onset of a LOCA and the containment air temperature is an approximation for the sump water temperature.

Containment Air Temperature = 271°F (Ref. 9.8, Table 4-2) = Sump Water Temperature The enthalpy information is obtained from Reference 9.50 hf, =239.97@271°F hf = 180.17 @ 212°F hfg = 970.3 @ 212°F Substituting the values in the above equation yields:

FF = 239.97 - 180.17 = 0.0616 or 6.16% which is less than the 10% value used in the analysis 970.3 6.13 Dose Adjustment for Aerosol Exclusion Area Boundary:

Containment leakage TEDE dose = 2.4213 rem TEDE (Sl5OCS75CLOl.Oo)

Containment leakage thyroid = 51.034 rem (Sl50CS75CL01.Oo)

Per RG 1.183 Footnote 7, the thyroid dose may be multiplied by 0.03 to obtain the thyroid TEDE equivalent dose:

Thyroid TEDE equivalent = 51.034 rem x 0.03 = 1.53102 rem TEDE Containment leakage whole body (WB) dose = 0.26625 rem (S150CS75CL01.Oo)

Containment leakage dose contribution to thyroid and WB

= 1.53102 rem TEDE + 0.26625 rem = 1.7973 rem TEDE Containment leakage TEDE dose from aerosols I Nuclear Common Revision 9 1 Nuclear Common Revision 9

I l CALCULATION CONTINUATION SHEET l SHEET 42 of 78 CALC. NO.: S-C-ZZ-MDC-1945

REFERENCE:

LCR S03-05 G. Patel/NUCORE, ORIGINATOR, DATE REV:

02/24/04 0

M Drucker/NUCORE, REVIEWERIVERIFIER, DATE 02/26/04

= 2.4213 rem TEDE - 1.7973 rem TEDE = 0.624 rem TEDE Plant-specific aerosol dose = 0.624 rem TEDE x 1.1 = 0.6864 rem TEDE Plant-specific containment leakage dose

= 1.7973 rem TEDE + 0.6864 TEDE = 2.484 rem TEDE, which is added to other post-LOCA dose contributions in Section 7.0 Low Population Zone:

Containment leakage TEDE dose = 0.45586 rem TEDE (SI5OCS75CLO1.Oo)

Containment leakage thyroid = 9.3066 rem (SI5OCS75CLO1.Oo)

Per RG 1.183 Footnote 7, the thyroid dose may be multiplied by 0.03 to obtain the thyroid TEDE equivalent dose:

Thyroid TEDE equivalent = 9.3066 rem x 0.03 = 0.2792 rem TEDE Containment leakage whole body (WB) dose = 0.066735 rem (SI50CS75CLO1.Oo)

Containment leakage dose contribution to thyroid and WB

= 0.2792 rem TEDE + 0.066735 rem = 0.34594 rem TEDE Containment leakage TEDE dose from aerosols

= 0.45586 rem TEDE - 0.34594 rem TEDE = 0.10992 rem TEDE Plant-specific aerosol dose = 0.10992 rem TEDE x 1.= 0.1209 rem TEDE Plant-specific containment leakage dose

= 0.34594 rem TEDE + 0.1209 TEDE = 0.4668 rem TEDE, which is added to other post-LOCA dose contributions in Section 7.0 Control Room:

Containment leakage TEDE dose = 0.70111 rem TEDE (S150CS75CLOI.Oo)

Containment leakage thyroid = 14.389 rem (SlIOCS75CLOI.Oo)

Per RG 1.183 Footnote 7, the thyroid dose may be multiplied by 0.03 to obtain the thyroid TEDE equivalent dose:

Thyroid TEDE equivalent = 14.389 rem x 0.03 = 0.4317 rem TEDE Containment leakage whole body (WB) dose = 0.10983 rem (S150CS75CLOI.Oo)

Containment leakage dose contribution to thyroid and WB

= 0.4317 rem TEDE + 0.10983 rem = 0.54153 rem TEDE Containment leakage TEDE dose from aerosols

= 0.70111 rem TEDE- 0.54153 rem TEDE = 0.15958 rem TEDE Plant-specific aerosol dose = 0.15958 rem TEDE x 1.1 = 0.17554 rem TEDE Plant-specific containment leakage dose

= 0.54153 rem TEDE + 0.17554 TEDE = 0.7171 rem TEDE, which is added to other post-LOCA dose contributions in Section 7.0 Nucl ar C

m monReviion I Nuclear Common Revision 9 1

I CALCULATION CONTINUATION SHEET ISHEET 43 of 78 CALC. NO.: S-C-ZZ-MDC-1945

REFERENCE:

LCR S03-05 G. Patel/NUCORE, ORIGINATOR, DATE REV:

02/24/04 0

M Drucker/NUCORE, REVIEWERIVERIFIER, DATE 02/26/04 7.0 RESULTS

SUMMARY

7.1 Post-LOCA EAB, LPZ, and CR doses at the Salem Nuclear Generating Station without the interruption of containment spray are summarized as follows:

Post-LOCA Post-LOCA TEDE Dose (Rem)

Activity Release Receptor Location Path Control Room EAB LPZ Containment Leakage 6.87E-01 2.27E+00 4.32E-01 (occurs @

0.5 hr)

ESF Leakage 2.79E+00 9.87E-01 7.45E-01 (occurs @

0.4 hr)

Containment Relief Line 3.99E-04 3.34E-05 5.67E-06 Release (occurs @

0.0 hr)

Back-leakage to RWST Negligible Negligible Negligible Containment Shine Negligible Negligible Negligible External Cloud 5.80E-02 N/A N/A CR Filter Shine 2.64E-01 N/A N/A Total 3.80E+00 3.26E+00 1.18E+00 Allowable TEDE Limit 5.OOE+00 2.50E+01 2.50E+01 RADTRAD Computer Run No.

Containment Leakage S150CS75CLO SCS75CL OO S 150CS75CLOO ESF Leakage S75ESF1GPM S75ESF1GPM S75ESF1GPM Cont. Relief Line Release SPURGEOO SPURGEOO SPURGEOO 7.2 Post-LOCA EAB, LPZ, and CR doses at the Salem Nuclear Generating Station with the 10 minutes interruption of containment spray are summarized as follows:

I Nuclear Common Revision 9

I CALCULATION CONTINUATION SHEET ISHEET 44 of 78 CALC. NO.: S-C-ZZ-MDC-1945

REFERENCE:

LCR S03-05 G. Patel/NUCORE, ORIGINATOR, DATE REV:

02/24/04 0

M Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/26/04 Post-LOCA Post-LOCA TEDE Dose (Rem)

Activity Release Receptor Location Path Control Room EAB LPZ Containment Leakage 7.17E-01 2.48E+00 4.67E-01 (occurs @

0.5 hr)

ESF Leakage 2.79E+00 9.87E-01 7.45E-0l (occurs @

0.4 hr)

Containment Relief Line 3.99E-04 3.34E-05 5.67E-06 Release (occurs @ 0.0 hr)

Back-leakage to RWST Negligible Negligible Negligible Containment Shine Negligible Negligible Negligible External Cloud 5.90E-02 N/A N/A CR Filter Shine 2.64E-01 N/A N/A Total 3.83E+00 3.47E+00 1.21E+00 Allowable TEDE Limit 5.OOE+O0 2.50E+01 2.50E+0l RADTRAD Computer Run No.

Containment Leakage Si 50CS75CLO1 Si 50CS75CLO1 S 150CS75CLO1 ESF Leakage S75ESF1 GPM S75ESF1 GPM S75ESF1 GPM Cont. Relief Line Release SPURGEOO SPURGEOO SPURGEOO

8.0 CONCLUSION

S:

The post-LOCA results presented in Section 7.2 indicate that the EAB, LPZ, and CR doses due to a LOCA are within their allowable limits.

Nucl ar C

m monReviion I Nuclear Common Revision 9 1

CALCULATION CONTINUATION SHEET ISHEET 45 of 78 CALC. NO.: S-C-ZZ-MDC-1945

REFERENCE:

LCR S03-05 G. Patel/NUCORE, ORIGINATOR, DATE IREV:

02/24/04 0

l l

M Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/26/04

9.0 REFERENCES

1.

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

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

3.

Westinghouse Calculation No. RSAC-PSE-800, 04/26/93, Source Term for Salem Margin Recovery

4.

Critical Software Package Identification No. A-0-ZZ-MCS-0225, Rev. 0, RADTRAD Computer Code, Version 3.02

5.

SNGS Calculation No. S-C-ZZ-MDC-1912, Rev 0, Control Room x/Qs Using ARCON96 Code -

Equipment Hatch & Plant Vent, and FHB Rollup Door Releases

6.

SNGS Technical Specifications:

6.1 Specification 1.25, Salem Unit 1/Unit 2 Rated Thermal Power 6.2 Specification Surveillance Requirement 4.7.6.l.d.1, Salem Unit 1/Unit 2 CREACS Ventilation Flow Rate 6.3 Specification Surveillance Requirement 4.7.6.l.d.3, Salem Unit 1/Unit 2 CREACS Design Makeup Flow Rate 6.4 Specification Limited Condition For Operation (LCO) 3.6.1.2.a, Salem Unit 1/Unit 2 Containment Leakage 6.5 Specification Surveillance Requirement 4.7.6.1.b.3, Salem Unit 1/Unit 2 CREACS Methyl Iodide Penetration 6.6 Specification Surveillance Requirement 4.7.6.1.e, Salem Unit 1/Unit 2 HEPA Filter DOP 6.7 Specification LCO 3.6.2.3, Salem Unit 1/Unit 2 Containment Cooling System 6.8 Not Used 6.9 Specification Table 3.3-5, Salem Unit 1/Unit 2 Engineered Safety Features Response Times 6.10 Specification 5.2.1, Salem Unit I/Unit 2 Containment Configuration 6.11 Specification 5.4.2, Salem Unit 1/Unit 2 Reactor Coolant System Volume 6.12 Specification LCO 3.4.8 (Salem Unit 1) and LCO 4.4.9 (Salem Unit 2), Reactor Coolant Specific Activity 6.13 Specification 1.10, Salem Unit 1/Unit 2 Dose Equivalent I-1 31

7.

SNGS Procedure No. LRT-Voll-MAN, Rev 2, Primary Containment Leak Rate Test - Program Manual

8.

Westinghouse WCAP-7952, Iodine Removal By Spray In The Salem Station Containment, August 1972

9.

U. S. NRC Standard Review Plan 6.5.2, Rev 2, Containment Spray As A Fission Product Cleanup System

10.

Vendor Technical Document No. 321035, Rev 3, Accident X/Q Values at the Salem Generating Station Control Room Fresh Air Intakes, Exclusion Area Boundary and Low Population Zone

11.

Westinghouse Letter No. PSE-96-535, Dated 02/05/1996, RWST Drain Down and Cold Leg Recirculation Radiological Consequences Clarification I Nuclear Common Revision 9 1 Nuclear Common Revision 9 I

CALCULATION CONTINUATION SHEET SHEET 46 of 78 CALC. NO.: S-C-ZZ-MDC-1945

REFERENCE:

LCR S03-05 G. PateVNUCORE, ORIGINATOR, DATE IREV:

02/24/04 0

l l

M DruckerfNUCORE, REVIEWER/VERIFIER, DATE 02t26/04

12.

SNGS Calculation No. S-C-VAR-NZZ-0020, Rev 0, UFSAR Chapter 15 DB/LB Accident Analysis Input Assumptions

13.

SNGS Engineering Evaluation No. S-C-ZZ-MEE-1797, Rev 0, ESF System Leakage to the Auxiliary Building & RWST

14.

SNGS Calculation No. S-C-CBV-MDC-1637, Rev 1, Containment Fan Cooler Unit Design Basis Capacity

15.

Westinghouse Calculation No. CN-CRA-93-206, Rev 0, Salem Post-LOCA Sump DF Limit

16.

SNGS Calculation No. S-C-VAR-MDC-1 575, Rev 1, Post-LOCA Recirculation ECCS Airborne Leakage Outside Containment

17.

Vendor Technical Document No. 127819, Rev 5, Auxiliary Building Pump Room Coolers

18.

CD P534 of Design Change Package (DCP) No. 1EC-3505, Rev 7, Package No. 1, Control Area Air Conditioning System Upgrade

19.

SNGS Duct Detail Mechanical Drawings:

a.

252220, Rev 2, No 1 Unit - Containment Bldg Above EL 130'-0" Sections & Details

b.

252260, Rev 1, No 2 Unit - Containment Bldg Above EL 130'-0" Sections & Details

20.

SNGS Architectural Drawings:

a.

207087, Sheet 1, Rev 7, No. 1 & 2 Units - Auxiliary Building Control Area - Floor Plan EL 122'-0"

b.

207089, Sheet 1, Rev 6, No. 1 & 2 Units - Auxiliary Building Control Area - Wall Sections &

Details

c.

207090, Sheet 1, Rev 3, No. 1 & 2 Units - Auxiliary Building Control Area - Wall Sections &

Details, Sheet 2

21.

SNGS Mechanical Arrangement Drawings:

a.

210488, Sheet 1, Rev 6, No. 1 Unit - Auxiliary Building Control Area Air Conditioning Equipment Room EL 122'-0"

a.

223540, Sheet 1, Rev 6, No. 2 Unit - Auxiliary Building Control Area Air Conditioning Equipment Room EL 122'-0"

b.

223299, Sheet 1, Rev 4, Yard - Yard Service Water Piping To Turbine Area

22.

SNGS Architectural Drawings:

a.

207082, Sheet 1, Rev 23, No. 1 Unit - Auxiliary Building Floor Plan EL 122'-0"

b.

207083, Sheet 1, Rev 19, No. 2 Unit - Auxiliary Building Floor Plan EL 122'-0"

23.

SNGS Concrete Structural Drawings:

a.

201047, Sheet 1, Rev 18, No. 1 & 2 Units - Auxiliary Building Slab EL 122'-0" Between Column Lines AA - FF Nuclear Common Revision 9 l

l

-CALCULATION CONTINUATION SHEET SHEET 47 of 78 CALC. NO.: S-C-ZZ-MDC-1945

REFERENCE:

LCR S03-05 G. Patel/NUCORE, ORIGINATOR, DATE IREV:

02/24/04 0

l l

M Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/26/04

b.

201051, Sheet 1, Rev 8, No. 1 & 2 Units - Auxiliary Building Roof Slab EL 140'-0" Between Column Lines AA - FF

c.

201092, Sheet 1, Rev 24, No. 1 & 2 Units - Auxiliary Building Section X - X, Sheet 1

d.

211750, Sheet 1, Rev 7, No. 1 & 2 Units - Class 1 Tanks & Pipe Tunnel Foundation - Plans &

Sections, Sheet 1 24 SNGS Calculation No. S-C-CBV-MDC-1438, Rev 0, Air Volume Released Through Containment Isolation Valves VC5 and VC6 for LOCA Occurring During Operation of the Pressure Relief System 25 SNGS Mechanical P&IDs:

a.

205248, Sheet 2, Rev 43, No. 1 Unit - Aux Bldg Control Area Air Conditioning & Ventilation

b.

205348, Sheet 2, Rev 34, No. 2 Unit - Aux Bldg Control Area Air Conditioning & Ventilation 26 USNRC, "Laboratory Testing of Nuclear-Grade Activated Charcoal", NRC Generic Letter 99-02, June 3, 1999 27 Federal Guidance Report 11, EPA-520/1-88-020, Environmental Protection Agency

28.

SNGS General Mechanical Arrangement Drawings:

a.

204804, Sheet 1, Rev 8, No. 1 & 2 Units - Auxiliary Building Reactor Containment, Fuel Building EL 100'

b.

204805, Sheet 1, Rev 6, No. I & 2 Units-Auxiliary Building EL 84' Reactor Cont. 78' & 81' Fuel Handling Area EL 85' & 89'-6"

c.

204807, Sheet 1, Rev 2, No. 1 & 2 Units - Auxiliary Building & Reactor Cont. EL 45' & 55'

d.

204808, Sheet 1, Rev 1, No. 1 & 2 Units - Auxiliary Building & Reactor Containment Section A-A

e.

204809, Sheet 1, Rev 6, No. 1 & 2 Units - Auxiliary Building & Reactor Containment Sections B-B & C-C

f.

204803, Sheet 1, Rev 11, No. 1 & 2 Units - Auxiliary Building EL 122' Reactor Containment &

Fuel Handling Area EL 130'

29.

SNGS Mechanical P&IDs:

a.

205237, Sheet 1, Rev 42, No. 1 Unit - Auxiliary Building Ventilation

b.

205237, Sheet 2, Rev 30, No. I Unit - Auxiliary Building Ventilation

c.

205237, Sheet 3, Rev 30, No. 1 Unit - Auxiliary Building Ventilation

d.

205337, Sheet 1, Rev 36, No. 2 Unit - Auxiliary Building Ventilation

e.

205337, Sheet 2, Rev 22, No. 2 Unit - Auxiliary Building Ventilation

f.

205337, Sheet 3, Rev 24, No. 2 Unit - Auxiliary Building Ventilation

30.

SNGS Architectural Drawings:

a.

207075, Sheet 1, Rev 12, No. 1 & 2 Units - Auxiliary Building Floor Plan EL 45'-0" & 55'-0" I Nuclear Common Revision 9 1

I CALCULATION CONTINUATION SHEET ISHEET 48 of 78 CALC. NO.: S-C-ZZ-MDC-1945

REFERENCE:

LCR S03-05 G. PatelINUCORE, ORIGINATOR, DATE IREV:

02/24/04 0

l l

M Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/26/04

b.

207078, Sheet 1, Rev 21, No. 1 Unit - Auxiliary Building Floor Plan EL 78'-0" & 84'-0"

c.

207079, Sheet 1, Rev 20, No. 2 Unit - Auxiliary Building Floor Plan EL 78'-0" & 84'-0"

31.

SNGS Concrete Structural Drawing No. 201030, Sheet 1, Rev 6, No. I & 2 Units -Auxiliary Building Residual Heat Removal Pit - Mat EL 45'-0"

32.

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

33.

Vendor Technical Document No. 326043, Control Room Envelope Inleakage Testing At Salem Nuclear Generating Station 2003

34.

SNGS Calculation S-C-ZZ-MDC-1987, Rev 1, Input Parameters for Salem AST Dose Calc

35.

ASME Steam Tables, Sixth Edition

36.

10 CFR 50.67, "Accident Source Term."

37.

SNGS Engineering Evaluation No. S-C-ZZ-MEE-1793, Rev 0, Post-LOCA pH Confirmation Calculation for Salem AST Phase-2 Project

38.

Specification No. S-C.-X400-SDS-0148-0, Structural Paint Specification for Reactor Containinents, Fuel Handling Buildings, and Auxiliary Building

39.

NUREG/CR-5950, Published December 1992, Iodine Evolution and pH Control

40.

SNGS Drawing No. 207095, Sheet 1, Rev 39, No 1 & 2 Units - Auxiliary Building Exterior & Interior Door Schedule

41.

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

42.

SNGS Calculation No. S-C-ZZ-MDC-1807, Rev 1, Loss of Coolant Accident Dose Consequences

43.

NUREG/CR-6331 (PNNL-10521), Rev 1, Atmospheric Relative Concentration in Building Wakes, May 1997.

44.

Critical Software Package No. A-O-ZZ-MCS-0224, Sheet 1, Rev 0, ARCON96 Computer Code Revision 0.

45.

Critical Software Package No. A-0-ZZ-MCS-0209, Sheet 1, Rev 0, MicroShield Computer Code Version 5.05.

46.

HCGS Drawing No. 239584, Rev. 2, "Location of Meteorological Tower and Road Plan and Details".

47.

SNGS Containment Ventilation System Flow Diagrams:

a.

205238, Sheet 1, Rev 35, No. I Unit Reactor Containment - Ventilation

b.

205238, Sheet 2, Rev 33, No. I Unit Reactor Containment - Ventilation

c.

205238, Sheet 3, Rev 31, No. 1 Unit Reactor Containment - Ventilation

d.

205338, Sheet 1, Rev 30, No. 2 Unit Reactor Containment - Ventilation

e.

205338, Sheet 2, Rev 27, No. 2 Unit Reactor Containment - Ventilation Nuclear Common Revision INuclear Common Revision 9 l

l CALCULATION CONTINUATION SHEET l SHEET 49 of 78 CALC. NO.: S-C-ZZ-MDC-1945

REFERENCE:

LCR S03-05 G. Patel/NUCORE, ORIGINATOR, DATE REV:

02/24/04 0

l l

M Drucker/NUCORE, REVIEWERNVERIFIER, DATE 02/26/04

f.

205338, Sheet 3, Rev 23, No. 2 Unit Reactor Containment - Ventilation

48.

SNGS Mechanical Arrangement Drawings:

a.

207635, Rev 7, No. 1 Unit - Reactor Containment Ventilation - Plan EL 130'-O" and Above

b.

207636, Rev 7, No. 1 Unit - Reactor Containment Ventilation - Plan Below EL 130'-0"

c.

207637, Rev 12, No. 1 Unit - Reactor Containment Ventilation - Sections

d.

207638, Rev 8, No. 2 Unit - Reactor Containment Ventilation - Plan EL 130'-0" and Above

e.

207639, Rev 9, No. 2 Unit - Reactor Containment Ventilation - Plan EL 130'-0"

50.

ASME Steam Tables, Sixth Edition I Nuclear Common Revision 9 1 Nucl ar om mo Re isio 9

I CALCULATION CONTINUATION SHEET ISHEET 50 of 78 CALC. NO.: S-C-ZZ-MDC-1945

REFERENCE:

LCR S03-05 G. PatellNUCORE, ORIGINATOR, DATE REV:

02/24/04 0

M Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/26/04 10.0 TABLES:

Table I Salem I & 2 Noble Gas & Iodine Normalized Core Inventory Core Core Normalized Inventory Power Core Isotope At 3600 MWt Normalizing Inventory (Ci)

Factor (Ci)

A B

C=AxB KR-85 1.100E+06 1.009 1.I11E+06 KR-85M 2.600E+07 1.009 2.623E+07 KR-87 4.700E+07 1.009 4.742E+07 KR-88 6.700E+07 1.009 6.760E+07 I-131 9.900E+07 1.009 9.988E+07 1-132 1.400E+08 1.009 1.412E+08 1-133 2.000E+08 1.009 2.018E+08 1-134 2.200E+08 1.009 2.220E+08 1-135 1.900E+08 1.009 1.917E+08 XE-133 2.000E+08 1.009 2.018E+08 XE-135 5.000E+07 1.009 5.044E+07 A From Reference 9.4, Table 2 B = (3459 MWt x 1.05)/3600 MWt = (3632/3600) = 1.009 Table 1A Comparison of RARTRAD Default & Westinghouse Core Inventory RADTRAD Default Value Westinghouse Core Inventory Core Isotope Inventory (Ci/MWt)

(Ci)

(Ci)

A B=Ax3600 C

1-131 2.540E+04 9.144E+07 9.900E+07 1-132 3.743E+04 1.347E+08 1.400E+08 1-133 5.370E+04 1.933E+08 2.000E+08 1-134 5.893E+04 2.121E+08 2.200E+08 1-135 5.063E+04 1.823E+08 1.900E+08 KR-85 1.960E+02 7.056E+05 1.100E+06 KR-85M 9.181E+03 3.305E+07 2.600E+07 KR-87 1.678E+04 6.041E+07 4.700E+07 KR-88 2.269E+04 8.168E+07 6.700E+07 XE-133 5.372E+04 1.934E+08 2.000E+08 XE-135 1.008E+04 3.629E+07 5.000E+07 A From RADTRAD Default NIF Pwr-def.nif C From Table I I Nuclear Common Revision 9 l

I CALCULATION CONTINUATION SHEET ISHEET 51 of 78 CALC. NO.: S-C-ZZ-MDC-1945

REFERENCE:

LCR S03-05 G. Patel/NUCORE, ORIGINATOR, DATE REV:

02/24/04 0

M Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/26/04 Table 2 Salem 1 & 2 Noble Gas & Iodine Nuclide Inventory File Normalized Core Core Core Thermal Inventory Isotope Inventory Power For RADTRAD (Ci)

(MWt)

NIF (CilMWt)

A B

C=A/B KR-85 1.110E+06 3632

.3056E+03 KR-85M 2.623E+07 3632

.7222E+04 KR-87 4.742E+07 3632

.1306E+05 KR-88 6.760E+07 3632

.1861E+05 I-131 9.988E+07 3632

.2750E+05 1-132 1.412E+08 3632

.3889E+05 I-133 2.018E+08 3632

.5556E+05 1-134 2.220E+08 3632

.611 1E+05 1-135 1.917E+08 3632

.5278E+05 XE-133 2.018E+08 3632

.5556E+05 XE-135 5.044E+07 3632

.1389E+05 A From Table I B = (3459 MWt x 1.05) = 3632 MWt l Nuclear Common Revision 9 Nuclear Common Revision 9

CALCULATION CONTINUATION SHEET ISHEET 52 of 78 CALC. NO.: S-C-ZZ-MDC-1945

REFERENCE:

LCR S03-05 G. Patel/NUCORE, ORIGINATOR, DATE REV:

02/24/04 0

M Drucker/NUCORE, REVIEWERIVERIFIER, DATE 02/26/04 Table 3 Salem I & 2 Core Inventory for RADTRAD NIF Salem_def Core Core Inventory For Inventory For Isotope RADTRAD Isotope RADTRAD NIF NIF (CitMWt)

(Ci/MWt)

A A

CO-58 2.553E+02 TE-131M 3.707E+03 CO-60 1.953E+02 TE-132 3.690E+04 KR-85*

3.056E+02 1-131*

2.750E+04 KR-85M*

7.222E+03 I-132*

3.889E+04 KR-87*

1.306E+04 I-133*

5.556E+04 KR-88*

1.861E+04 1-134*

6.I1 lE+04 RB-86 1.496E+01 I-135*

5.278E+04 SR-89 2.844E+04 XE-133*

5.556E+04 SR-90 1.535E+03 XE-135*

1.389E+04 SR-91 3.656E+04 CS-134 3.425E+03 SR-92 3.805E+04 CS-136 1.042E+03 Y-90 1.647E+03 CS-137 1.915E+03 Y-91 3.465E+04 BA-139 4.976E+04 Y-92 3.819E+04 BA-140 4.924E+04 Y-93 4.320E+04 LA-140 5.032E+04 ZR-95 4.377E+04 LA-141 4.615E+04 ZR-97 4.562E+04 LA-142 4.449E+04 NB-95 4.13 8E+04 CE-141 4.476E+04 MO-99 4.830E+04 CE-143 4.352E+04 TC-99M 4.169E+04 CE-144 2.697E+04 RU-103 3.598E+04 PR-143 4.273E+04 RU-105 2.340E+04 ND-147 1.91 IE+04 RU-106 8.175E+03 NP-239 5.120E+05 RH-105 1.621E+04 PU-238 2.902E+01 SB-127 2.208E+03 PU-239 6.545E+00 SB-129 7.820E+03 PU-240 8.254E+00 TE-127 2.132E+03 PU-241 1.390E+03 TE-127M 2.823E+02 AM-241 9.181 E-0 1 TE-129 7.341E+03 CM-242 3.514E+02 TE-129M 1.935E+03 CM-244 2.056E+01

• Noble Gas & Iodine Inventory From Table 2 A = Aerosol Inventory From RADTRAD Default NIF Pwr defnif I Nuclear Common Revision 9 1 Nuclear Common Revision 9

I CALCULATION CONTINUATION SHEET SHEET 53 of 78 CALC NO.: S-C-ZZ-MDC-1945

REFERENCE:

LCR S03-05 G. PatelVNUCORE, ORIGINATOR, DATE IREV:

02/24/04 0

M Drucker/NUCORE, REVIEWERtVERIFIER, DATE 02/26/04 Table 4 Iodine Isotopic Thyroid Inhalation Dose Conversion Factors Iodine Conversion Iodine Dose Factor Dose Isotope Conversion Conversion Factor Factor (Sv/Bq)

(rem/Ci per Sv/Bq)

(rem/Ci)

A B

C=AxB 1-131 2.92E-07 3.70E+12 1.08E+06 1-132 1.74E-09 3.70E+12 6.44E+03 I-133 4.86E-08 3.70E+12 1.80E+05 1-134 2.88E-10 3.70E+12 1.07E+03 1-135 8.46E-09 3.70E+12 3.13E+04 A From Reference 9.27, Page 136 Table 5 Iodine Scaling Factor to Obtain RCS Tech Spec Activity of 1.0 JCi/g DE 1-131 1% Failed Fuel Iodine Iodine Dose Product Isotope Activity Conversion Concentration Factor (u1Cig)

(reml/Ci)

(jiCi.rern1Ci.g)

A B

(A x B) 1-131 2.80E+00 1.08E+06 3.02E+06 1-132 2.80E+00 6.44E+03 1.80E+04 I-133 4.20E+00 1.80E+05 7.55E+05 1-134 5.70E-01 1.07E+03 6.08E+02 1-135 2.30E+00 3.13E+04 7.20E+04 Total 3.87E+06 A From Reference 9.3, Table 4 I-131 DE Based on 1% FF Iodine Concentration

= Total Iodine Activity divided by 1-131 DCF 3.58E+00 Iodine Scaling Factor to Obtain 1.0 piCig DE 1-131

= (1.0 pCi/g DE 1-131) / (3.58 pCi/g DE 1-131) 2.791E-01 I Nuclear C-omm-on Revision 9 1 Nuclear Common Revision 9

I CALCULATION CONTINUATION SHEET ISHEET 54 of 78 CALC. NO.: S-C-ZZ-MDC-1945

REFERENCE:

LCR S03-05 G. Patel/NUCORE, ORIGINATOR, DATE IREV:

02/24/04 0

M Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/26/04 Table 6 Iodine Concentration Based on 1.0 pCi/g DE 1-131 1% Failed Fuel Iodine 1.0 pCi/g DE 1-131 Iodine Scaling Iodine Spike Isotope Activity Factor Activity Concentration 1.0,uCi/g Concentration

([LCVg)

DE 1-131 (6Cilg)

A B

C=AxB I-131 2.80E+00 2.791E-01 7.8 1E-01 1-132 2.80E+00 2.791E-01 7.81E-01 1-133 4.20E+00 2.791E-01 1.17E+00 1-134 5.70E-01 2.791E-01 1.59E-01 1-135 2.30E+00 2.791E-01 6.42E-01 A From Reference 9.3, Table 4 B Scaling Factor Based on 1.0 pCi/g DE 1-131 From Table 5 I Nuclear Common Revision 9 1 Nuclear Common Revision 9

I CALCULATION CONTINUATION SHEET ISHEET 55 of 78 CALC. NO.: S-C-ZZ-MDC-1945

REFERENCE:

LCR S03-05 G. Patel/NUCORE, ORIGINATOR, DATE IREV:

02/24/04 0

M Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/26/04 Table 7 RCS Activity Released in Containment From Vacuum Relief Line Release Tech Spec Reactor Total Reactor Coolant Coolant Reactor Isotope Activity Mass Coolant Concentration Activity (jtCilg)

(g)

(Ci)

A B

AxBxlE-6 Kr-83m 4.OOE-01 2.511E+08 11.OOE+02 Kr-85 8.20E+00 2.51 E+08 2.05E+03 Kr-85m 1.70E+00 2.511E+08 4.26E+02 Kr-87 1.00E+00 2.51E+08 2.51E+02 Kr-88 3.00E+00 2.5 IE+08 7.52E+02 Xe-131m 2.10E+00 2.51E+08 5.26E+02 Xe-133 2.60E+02 2.5 IE+08 6.52E+04 Xe-133m 1.70E+01 2.5 1E+08 4.26E+03 Xe-135 8.50E+00 2.51E+08 2.13E+03 Xe-135m 4.90E-01 2.51E+08 1.23E+02 Xe-138 6.10E-01 2.511E+08 1.53E+02 I-131 7.81E-01 2.51E+08 1.96E+02 I-132 7.81 E-0 1 2.51 E+08 1.96E+02 I-133 1.17E+00 2.51E+08 2.94E+02 I-134 1.59E-01 2.51E+08 3.99E+01 I-135 6.42E-01 2.51 E+08 1.61 E+02 A-l% FF Noble gas concentrations from Reference 9.3, Table 4 A-Tech Spec Iodine concentrations from Table 6 B Reactor coolant mass from Section 6.7 N u c l a r C m m o nR e v i i o n I Nuclear Common Revision 9 1

I CALCULATION CONTINUATION SHEET ISHEET 56 of 78 CALC. NO.: S-C-ZZ-MDC-1945

REFERENCE:

LCR S03-05 G. PateVNUCORE, ORIGINATOR, DATE REV:

02/24/04 0

M Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/26/04 Table 8 RDATRAD NIF for Vacuum Relief Line Release Total Core RADTRAD Reactor Thermal NIF For Isotope Coolant Power Vacuum Activity Level Relief (Ci)

(MW,)

(Ci/MW,)

A B

Kr-83m 1.OOE+02 3632

.2760E-01 Kr-85 2.05E+03 3632

.5658E+00 Kr-85m 4.26E+02 3632

.1317311+00 Kr-87 2.51E+02 3632

.6900E-01 Kr-88 7.52E+02 3632

.2070E+00 Xe-131m 5.26E+02 3632

.1449E+00 Xe-133 6.52E+04 3632

.1794E+02 Xe-133m 4.26E+03 3632

.1173E+01 Xe-135 2.13E+03 3632

.5865E+00 Xe-135m 1.23E+02 3632

.3381E-01 Xe-138 1.53E+02 3632

.4209E-01 I-131 1.96E+02 3632

.5392E-01 I-132 1.96E+02 3632

.5392E-01 1-133 2.94E+02 3632

.8088E-01 1-134 3.99E+01 3632

.1098E-01 1-135 1.61E+02 3632

.4429E-01 A From Table 7 B = 3459 MWt x 1.05 = 3632 MWt l Nuclear Common Revision 9 l Nuclear Common Revision 9 I

CALCULATION CONTINUATION SHEET SHEET 57 of 78 CALC. NO.: S-C-ZZ-MDC-1945

REFERENCE:

LCR S03-05 G. Patel/NUCORE, ORIGINATOR, DATE REV:

02/24/04 0

M Drucker/NUCORE, REVIEWERIVERIFIER, DATE 02126/04 Table 9 Conversion of Iodine Activity Into Iodine Atoms Sprayed Region @ 0.5 hr Iodine Isotopic Isotope Inventory Inventory (Atoms Per Iodine (Curie)

(Atoms)

Curie)

Fraction A

B Ci=Bi/Ai Di=Bi/ZB 1-131 1.069E+06 3.963E+22 3.708E+1 6 7.63 1 E-0 1 1-132 1.51 IE+06 6.680E+20 4.420E+14 1.286E-02 1-133 2.159E+06 8.63 1E+21 3.997E+15 1.662E-01 I-134 2.375E+06 4.00 1E+20 1.685E+14 7.704E-03 1-135 2.051E+06 2.606E+21 1.270E+15 5.017E-02 Total 5.193E+22 1.OOOE+00 A & B From RADTRAD Run S150CS9OCLO1 output file @ 0.5 hr from Sprayed Region Compartment Nuclide Inventory Table 10 Post-LOCA Elemental Iodine Isotopic Activity In Sump ( 4.0 Hrs Iodine Fraction Elemental I Elemental Iodine Isotopic Isotope Atoms Per Of Iodine Atoms In Atoms Activity Curie Sump Sump In Sump 4.0 hr At 4.0 Hr At 4.0 Hr (Ci)

A B

C Di =Bi

• C Ei=Di/Ai 1-131 3.708E+16 7.631E-01 9.423E+22 7.190E+22 1.939E+06 1-132 4.420E+14 1.286E-02 1.212E+21 2.742E+06 1-133 3.997E+15 1.662E-01 1.566E+22 3.918E+06 1-134 1.685E+14 7.704E-03 7.259E+20 4.309E+06 1-135 1.270E+15 5.017E-02 4.727E+21 3.722E+06 Total Iodine Sump Atoms/Activity 9.423E+22 1.663E+07 A & B From Table 9 C From RADTRAD S150CS9OCLO.oO output file @ 4.0 hrs from under Sprayed Region Transport Group inventory for elemental iodine atoms in sump I Nuclear Common Revision 9 1 I Nu l a o

m nR v s o

I CALCULATION CONTINUATION SHEET ISHEET 58 of 78 CALC NO.: S-C-ZZ-MDC-1945

REFERENCE:

LCR S03-05 G. Patel/NUCORE, ORIGINATOR, DATE IREV:

02/24/04 0

M Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/26/04 Table 11 Sump Elemental Iodine Activity In Fraction of Core Iodine Activity Isotope Core Sump Activity Iodine Expected Expected Calculated Elemental Iodine Activity Iodine Elemental RADTRAD Fraction of Iodine Core Activity (Ci)

(Ci)

(Ci)

(Ci)

(Ci)

A B=0.4xA C=0.0485xB D

E=(3*C)/A I-131 9.988E+07 3.995E+07 1.938E+06 1.939E+06 5.824E-02 1-132 1.412E+08 5.650E+07 2.740E+06 2.742E+06 5.824E-02 I-133 2.018E+08 8.07 1E+07 3.914E+06 3.918E+06 5.825E-02 1-134 2.220E+08 8.878E+07 4.306E+06 4.309E+06 5.824E-02 1-135 1.9177E+08 7.668E+07 3.7199E+06 3.722E+06 5.824E-02 A From Table I D From Table 10 E represents the fraction of core iodine activity released into the ESF recirculation loop leakage source term, including a factor of 3 for conservatism.

Nu l a o m nR v s o I Nuclear Common Revision 9 1

I CALCULATION CONTINUATION SHEET I SHEET 59 of 78 CALC. NO.: S-C-ZZ-MDC-1945

REFERENCE:

LCR S03-05 G. Patel/NUCORE, ORIGINATOR, DATE REV:

02/24/04 0

M Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/26/04 Table 12 Relationship of Aerosol Mass and Activity CR Region @ 0.8 hr Aerosol Isotopic Isotope Activity Mass Mass Per Ci Aerosol (Curie)

(kg)

(kg/Ci)

Fraction A

Bi Ci = Bi /Ai Di = Bi/EB Co-58 1.088E-07 3.420E-15 3.145E-08 1.716E-06 Co-60 8.325E-08 7.364E-14 8.847E-07 3.695E-05 Rb-86 1.168E-06 1.436E-14 1.229E-08 7.204E-06 Sr-89 9.690E-05 3.336E-12 3.442E-08 1.674E-03 Sr-90 5.234E-06 3.837E-1 1 7.33 IE-06 1.925E-02 Sr-91 1.126E-04 3.106E-14 2.759E-10 1.558E-05 Sr-92 9.077E-05 7.22 1E-15 7.956E-1 I 3.623E-06 Y-90 5.532E-08 1.017E-16 1.838E-09 5.102E-08 Y-91 1.181E-06 4.815E-14 4.077E-08 2.416E-05 Y-92 9.906E-07 1.030E-16 1.039E-10 5.166E-08 Y-93 1.338E-06 4.012E-16 2.997E-10 2.013E-07 Zr-95 1.492E-06 6.943E-14 4.655E-08 3.484E-05 Zr-97 1.469E-06 7.684E-16 5.231E-10 3.856E-07 Nb-95 1.41 OE-06 3.605E-14 2.557E-08 1.809E-05 Mo-99 2.029E-05 4.230E-14 2.085E-09 2.122E-05 Tc-99m 1.513E-05 2.877E-15 1.902E-10 1.444E-06 Ru-103 1.532E-05 4.747E-13 3.098E-08 2.382E-04 Ru-105 8.020E-06 1.193E-15 1.488E-10 5.986E-07 Ru-106 3.484E-06 1.04 1E-12 2.989E-07 5.225E-04 Rh-lOS 6.723E-06 7.965E-15 1.185E-09 3.996E-06 Sb-127 1.863E-05 6.975E-14 3.745E-09 3.500E-05 Sb-129 5.328E-05 9.474E-15 1.778E-10 4.754E-06 Te-127 1.639E-05 6.209E-15 3.789E-10 3.116E-06 Te-127m 2.406E-06 2.551 E-13 1.060E-07 1.280E-04 Te-129 2.718E-05 1.298E-15 4.775E-11 6.51 IE-07 Te-129m 1.648E-05 5.469E-13 3.3199E-08 2.744E-04 I Nuclear Common Revision 9 1 Nuclear Common Revision 9 I

CALCULATION CONTINUATION SHEET SHEET 60 of 78 CALC. NO.: S-C-ZZ-MDC-1945

REFERENCE:

LCR S03-05 G. Patel/NUCORE, ORIGINATOR, DATE REV:

02/24/04 0

M Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/26/04 Table 12 (Cont'd)

Relationship of Aerosol Mass and Activity CR Region @ 0.8 hr Aerosol Isotopic Isotope Activity Mass Mass Per Ci Aerosol (Curie)

(kg)

(kg/Ci)

Fraction A

Bi Ci=Bi/Ai Di=Bi/EB Te-13 Im 3.060E-05 3.837E-14 1.254E-09 1.925E-05 Te-132 3.107E-04 1.023E-12 3.294E-09 5.135E-04 Cs-134 2.680E-04 2.071 E-10 7.729E-07 1.03 9E-0 1 Cs-136 8.130E-05 1.109E-12 1.364E-08 5.566E-04 Cs-137 1.50E-04 1.72E-09 1.150E-05 8.643E-01 Ba-139 8.41E-05 5.14E-15 6.114E-11 2.579E-06 Ba-140 1.674E-04 2.286E-12 1.366E-08 1.147E-03 La-140 1.675E-06 3.014E-1 5 1.799E-09 1.5 12E-06 La-141 1.230E-06 2.175E-16 1.768E-10 1.091E-07 La-142 8.098E-07 5.657E-17 6.986E-11 2.838E-08 Ce-141 3.81 IE-06 1.338E-13 3.509E-08 6.71 IE-05 Ce-143 3.603E-06 5.425E-15 1.506E-09 2.722E-06 Ce-144 2.299E-06 7.20SE-13 3.1355E-07 3.617E-04 Pr-143 1.453E-06 2.157E-14 1.485E-08 1.082E-05 Nd-147 6.493E-07 8.026E-15 1.236E-08 4.027E-06 Np-239 4.29 lE-05 1.850E-13 4.310iE-09 9.280E-05 Pu-238 2.474E-09 1.445E-13 5.84 IE-05 7.25 IE-05 Pu-239 5.580E-10 8.977E-12 1.609E-02 4.504E-03 Pu-240 7.037E-10 3.088E-12 4.388E-03 1.549E-03 Pu-241 1.185E-07 1.150E-12 9.707E-06 5.772E-04 Am-241 3.131E-11 9.122E-15 2.914E-04 4.577E-06 Cm-242 1.198E-08 3.615E-15 3.017E-07 1.814E-06 Cm-244 7.01 IE-10 8.666E-15 1.236E-05 4.348E-06 Total 1.993E-09 I.OOOE+00 A & B From RADTRAD Run S150CS75CL01 output file @ 0.8 hr from Control Room Compartment Nuclide Inventory I Nuclear Common Revision 9 1 I Nu l a o m nR v s o

I CALCULATION CONTINUATION SHEET ISHEET 61 of 78 CALC. NO.: S-C-ZZ-MDC-1945

REFERENCE:

LCR S03-05 G. Patel/NUCORE, ORIGINATOR, DATE REV:

02/24/04 0

M Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/26/04 Table 13 Post-LOCA Total Aerosol Isotopic Activity On CR HEPA Filter 720 Hrs Post-LOCA Containment Leakage Aerosol Fraction Total Aerosol Isotopic Isotope Mass Per Ci of CR Filter Aerosol Mass Aerosol Activity Aerosol Aerosol Mass On CR Filter On CR Filter At 720 Hr At 720 Hr At 720 Hr (kg/Ci)

(kg)

(kg)

(Ci)

A Bi C

Di =Bi

• C Ei = Di / Ai Co-58 3.145E-08 1.716E-06 6.162E-07 1.058E-12 3.363E-05 Co-60 8.847E-07 3.695E-05 6.162E-07 2.277E-1 I 2.574E-05 Rb-86 1.229E-08 7.204E-06 6.162E-07 4.439E-12 3.612E-04 Sr-89 3.442E-08 1.674E-03 6.162E-07 1.031E-09 2.996E-02 Sr-90 7.33 IE-06 1.925E-02 6.162E-07 1.186E-08 1.6188E-03 Sr-91 2.759E-10 1.558E-05 6.162E-07 9.603E-12 3.481E-02 Sr-92 7.956E-11 3.623E-06 6.162E-07 2.233E-12 2.806E-02 Y-90 1.838E-09 5.102E-08 6.162E-07 3.144E-14 1.710E-05 Y-91 4.077E-08 2.416E-05 6.162E-07 1.489E-11 3.651E-04 Y-92 1.039E-10 5.166E-08 6.162E-07 3.183E-14 3.063E-04 Y-93 2.997E-10 2.013E-07 6.162E-07 1.240E-13 4.138E-04 Zr-95 4.655E-08 3.484E-05 6.162E-07 2.147E-11 4.612E-04 Zr-97 5.231E-10 3.856E-07 6.162E-07 2.376E-13 4.542E-04 Nb-95 2.557E-08 1.809E-05 6.162E-07 1.114E-11 4.358E-04 Mo-99 2.085E-09 2.122E-05 6.162E-07 1.308E-1 I 6.273E-03 Tc-99m 1.902E-10 1.444E-06 6.162E-07 8.896E-13 4.678E-03 Ru-103 3.098E-08 2.382E-04 6.162E-07 1.468E-10 4.737E-03 Ru-105 1.488E-10 5.986E-07 6.162E-07 3.689E-13 2.480E-03 Ru-106 2.989E-07 5.225E-04 6.162E-07 3.220E-10 1.077E-03 Rh-105 1.185E-09 3.996E-06 6.162E-07 2.463E-12 2.079E-03 Sb-127 3.745E-09 3.500E-05 6.162E-07 2.157E-1 1 5.759E-03 Sb-129 1.778E-10 4.754E-06 6.162E-07 2.929E-12 1.647E-02 Te-127 3.789E-10 3.116E-06 6.162E-07 1.920E-12 5.067E-03 Te-127m 1.060E-07 1.280E-04 6.162E-07 7.886E-1 1 7.438E-04 Te-129 4.775E-11 6.51 IE-07 6.162E-07 4.012E-13 8.402E-03 Te-129m 3.3199E-08 2.744E-04 6.162E-07 1.69 IE-10 5.094E-03 I Nuclear Common Revision 9 1 Nuclear Common Revision 9

CALCULATION CONTINUATION SHEET SHEET 62 of 78 CALC. NO.: S-C-ZZ-MDC-1945

REFERENCE:

LCR S03-05 G. Patel/NUCORE, ORIGINATOR, DATE REV:

02/24/04 0

M Drucker/NUCORE, REVIEWERIVERIFIER, DATE 02/26/04 Table 13 (Cont'd)

Post-LOCA Total Aerosol Isotopic Activity On CR HEPA Filter @ 720 Hrs Post-LOCA Containment Leakage Aerosol Fraction Total Aerosol Isotopic Isotope Mass Per Ci of CR Filter Aerosol Mass Aerosol Activity Aerosol Aerosol Mass On CR Filter On CR Filter At 720 Hr At 720 Hr At 720 Hr (kg/Ci)

(kg)

(kg)

(Ci)

A Bi C

Di=Bi* C Ei=Di/Ai Te-131m 1.254E-09 1.925E-05 6.162E-07 1.186E-11 9.460E-03 Te-132 3.294E-09 5.135E-04 6.162E-07 3.164E-10 9.606E-02 Cs-134 7.729E-07 1.039E-01 6.162E-07 6.403E-08 8.285E-02 Cs-136 1.364E-08 5.566E-04 6.162E-07 3.430E-10 2.514E-02 Cs-137 1.150E-05 8.643E-01 6.162E-07 5.326E-07 4.632E-02 Ba-139 6.114E-11 2.579E-06 6.162E-07 1.589E-12 2.600E-02 Ba-140 1.366E-08 1.147E-03 6.162E-07 7.069E-10 5.175E-02 La-140 1.799E-09 1.512E-06 6.162E-07 9.3181E-13 5.179E-04 La-141 1.768E-10 1.091E-07 6.162E-07 6.725E-14 3.803E-04 La-142 6.986E-1 I 2.838E-08 6.162E-07 1.749E-14 2.504E-04 Ce-141 3.509E-08 6.71 lE-05 6.162E-07 4.135E-1 1 1.178E-03 Ce-143 1.506E-09 2.722E-06 6.162E-07 1.677E-12 1.1 14E-03 Ce-144 3.135E-07 3.617E-04 6.162E-07 2.228E-10 7.1083E-04 Pr-143 1.485E-08 1.082E-05 6.162E-07 6.670E-12 4.492E-04 Nd-147 1.236E-08 4.027E-06 6.162E-07 2.481E-12 2.007E-04 Np-239 4.31 OE-09 9.280E-05 6.162E-07 5.7181E-11 1.327E-02 Pu-238 5.841E-05 7.251E-05 6.162E-07 4.468E-1 I 7.649E-07 Pu-239 1.609E-02 4.504E-03 6.162E-07 2.775E-09 1.725E-07 Pu-240 4.388E-03 1.549E-03 6.162E-07 9.548E-10 2.176E-07 Pu-241 9.707E-06 5.772E-04 6.162E-07 3.557E-10 3.664E-05 Am-241 2.914E-04 4.577E-06 6.162E-07 2.820E-12 9.680E-09 Cm-242 3.017E-07 1.814E-06 6.162E-07 1.118E-12 3.704E-06 Cm-244 1.236E-05 4.348E-06 6.162E-07 2.679E-12 2.168E-07 Total CR Aerosol Mass/Activity @ 720 hrs 6.162E-07 5.159E-01 A & B From Table 12 C From Section 6.9.2 I Nuclear Common Revision 9 1 I u l a o m nR v s o

I CALCULATION CONTINUATION SHEET ISHEET 63 of 78 CALC. NO.: S-C-ZZ-MDC-1945

REFERENCE:

LCR S03-05 G. Patel/NUCORE, ORIGINATOR, DATE REV:

02/24/04 0

M Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/26/04 Table 14 Post-LOCA Cont. Leakage Iodine Activity Deposited on CR Charcoal Filter Iodine Fraction Elemental &

Iodine Iodine Isotope Atoms Per Of Iodine Organic Iodine Atoms on Activity Curie Atoms On CR Charcoal CR Charcoal CR Charcoal Filter Filter 720 Hrs At 720 Hrs At 720 Hrs Ci A

B C

Di=Bi

• C Ei=Di/Ai 1-131 3.708E+16 7.631E-01 1.752E+16 1.337E+16 3.605E-01 I-132 4.420E+14 1.286E-02 2.253E+14 5.098E-01 I-133 3.997E+15 1.662E-01 2.91 1E+15 7.283E-01 1-134 1.685E+14 7.704E-03 1.349E+14 8.01OE-01 I-135 1.270E+15 5.017E-02 I

_8.788E+14 6.918E-01 Total Iodine Sump Atoms/Activity 1.752E+16 3.091E+00 A & B From Table 9 C From Section 6.9.1 I Nuclear Common Revision 9 1 I Nuclear Common Revision 9 I

I CALCULATION CONTINUATION SHEET ISHEET 64 of 78 CALC. NO.: S-C-ZZ-MDC-1945

REFERENCE:

LCR S03-05 G. Patel/NUCORE, ORIGINATOR, DATE REV:

02/24/04 0

M Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/26/04 Table 15 Containment Leakage Net Total Activity on CR Charcoal/HEPA Filter Isotope 0-720 Isotope 0-720 Co-58 3.363E-05 Te-131m 9.460E-03 Co-60 2.574E-05 Te-132 9.606E-02 Rb-86 3.612E-04 Cs-134 8.285E-02 Sr-89 2.996E-02 Cs-136 2.514E-02 Sr-90 1.618E-03 Cs-137 4.632E-02 Sr-91 3.481E-02 Ba-139 2.600E-02 Sr-92 2.806E-02 Ba-140 5.175E-02 Y-90 1.710E-05 La-140 5.179E-04 Y-91 3.651E-04 La-141 3.803E-04 Y-92 3.063E-04 La-142 2.504E-04 Y-93 4.138E-04 Ce-141 1.178E-03 Zr-95 4.612E-04 Ce-143 1.114E-03 Zr-97 4.542E-04 Ce-144 7.108E-04 Nb-95 4.358E-04 Pr-143 4.492E-04 Mo-99 6.273E-03 Nd-147 2.007E-04 Tc-99m 4.678E-03 Np-239 1.327E-02 Ru-103 4.737E-03 Pu-238 7.649E-07 Ru-105 2.480E-03 Pu-239 1.725E-07 Ru-106 1.077E-03 Pu-240 2.176E-07 Rh-105 2.079E-03 Pu-241 3.664E-05 Sb-127 5.759E-03 Am-241 9.680E-09 Sb-129 1.647E-02 Cm-242 3.704E-06 Te-127 5.067E-03 Cm-244 2.168E-07 Te-127m 7.438E-04 I-131 3.605E-01 Te-129 8.402E-03 1-132 5.098E-01 Te-129m 5.094E-03 1-133 7.283E-01 1-134 8.010E-01 1-135 6.918E-01 Aerosol Activity From Table 13 Iodine Activity From Table 14 I Nuclear Common Revision 9 1 Nu l a Co m nR v s o

I CALCULATION CONTINUATION SHEET SHEET 65 of 78 CALC. NO.: S-C-ZZ-MDC-1945

REFERENCE:

LCR S03-05 G. Patel/NUCORE, ORIGINATOR, DATE REV:

02/24/04 0

M Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/26/04 Table 16 Post-LOCA ESF Leakage Elemental & Organic Iodine Activity Deposited on CR Filter Iodine Fraction Elemental &

Iodine Iodine Isotope Atoms Per Of Iodine Organic Iodine Atoms on Activity Curie Atoms On CR Charcoal CR Charcoal CR Charcoal Filter Filter 720 Hrs At 720 Hrs At 720 Hrs Ci A

B C

Di =Bi

• C Ei=Di/Ai 1-131 3.708E+16 7.631E-01 9.813E+17 7.488E+17 2.019E+01 1-132 4.420E+14 1.286E-02 1.262E+16 2.856E+01 1-133 3.997E+15 1.662E-01 1.63 IE+17 4.080E+01 1-134 1.685E+14 7.704E-03 7.560E+15 4.487E+01 1-135 1.270E+15 5.017E-02 4.923E+16 3.876E+01 Total Iodine Sump Atoms/Activity 9.813E+17 1.732E+02 A & B From Table 9 C From Section 6.10.1 I Nuclear Common Revision 9 1 I Nuclear Common Revision 9 I

I CALCULATION CONTINUATION SHEET ISHEET 66 or 78 CALC.NO.: S-C-ZZ-MDC-1945

REFERENCE:

LCR S03-05 G. Patel/NUCORE, ORIGINATOR, DATE REV:

02/24/04 0

M Drucker/NUCORE, REVIEWERJVERIFIER, DATE 02/26/04 Table 17 720-hrs Post-LOCA Total Iodine & Aerosol Activity on CR CharcoallHEPA Filters (Ci)

Cont.

ESF Cont.

ESF Isotope Leakage Leakage Total Isotope Leakage Leakage Total A

B A+B A

B A+B Co-58 3.363E-05 0.000E+00 3.363E-05 Te-131m 9.460E-03 0.000E+00 9.460E-03 Co-60 2.574E-05 0.000E+00 2.574E-05 Te-132 9.606E-02 0.000E+00 9.606E-02 Rb-86 3.612E-04 0.000E+00 3.612E-04 Cs-134 8.285E-02 0.000E+00 8.285E-02 Sr-89 2.996E-02 0.000E+00 2.996E-02 Cs-136 2.514E-02 0.000E+00 2.514E-02 Sr-90 1.618E-03 0.000E+00 1.618E-03 Cs-137 4.632E-02 0.000E+00 4.632E-02 Sr-91 3.481IE-02 0.000E+00 3.481E-02 Ba-139 2.600E-02 0.OOOE+00 2.600E-02 Sr-92 2.806E-02 0.OOOE+00 2.806E-02 Ba-140 5.175E-02 0.000E+00 5.175E-02 Y-90 1.71 OE-05 0.OOOE+00 1.71 OE-05 La-140 5.179E-04 0.000E+00 5.179E-04 Y-91 3.65SIE-04 0.OOOE+00 3.651E-04 La-141 3.803E-04 0.000E+00 3.803E-04 Y-92 3.063E-04 0.000E+00 3.063E-04 La-142 2.504E-04 0.000E+00 2.504E-04 Y-93 4.138E-04 0.000E+00 4.138E-04 Ce-141 1.178E-03 0.OOOE+00 1.178E-03 Zr-95 4.612E-04 0.000E+00 4.612E-04 Ce-143 1.1 14E-03 0.000E+00 1.1 14E-03 Zr-97 4.542E-04 0.000E+00 4.542E-04 Ce-144 7.108E-04 O.OOOE+00 7.108E-04 Nb-95 4.358E-04 0.000E+00 4.358E-04 Pr-143 4.492E-04 0.000E+00 4.492E-04 Mo-99 6.273E-03 0.000E+00 6.273E-03 Nd-147 2.007E-04 0.000E+00 2.007E-04 Tc-99m 4.678E-03 0.000E+00 4.678E-03 Np-239 1.327E-02 0.000E+00 1.327E-02 Ru-103 4.737E-03 0.000E+00 4.737E-03 Pu-238 7.649E-07 0.000E+00 7.649E-07 Ru-105 2.480E-03 0.000E+00 2.480E-03 Pu-239 1.725E-07 0.000E+00 1.725E-07 Ru-106 1.077E-03 0.000E+00 1.077E-03 Pu-240 2.176E-07 0.000E+00 2.176E-07 Rh-105 2.079E-03 0.000E+00 2.079E-03 Pu-241 3.664E-05 0.000E+00 3.664E-05 Sb-127 5.759E-03 0.000E+00 5.759E-03 Am-241 9.680E-09 0.OOOE+00 9.680E-09 Sb-129 1.647E-02 O.OOOE+00 1.647E-02 Cm-242 3.704E-06 0.OOOE+00 3.704E-06 Te-127 5.067E-03 0.000E+00 5.067E-03 Cm-244 2.168E-07 0.000E+00 2.168E-07 Te-127m 7.438E-04 0.000E+00 7A38E-04 1-131 3.605E-01 2.019E+01 2.055E+01 Te-129 8.402E-03 0.OOOE+00 8.402E-03 1-132 5.098E-01 2.856E+01 2.907E+01 Te-129m 5.094E-03 0.000E+00 5.094E-03 1-133 7.283E-01 4.080E+01 4.153E+01 1-134 8.01OE-01 4.487E+01 4.567E+01 1-135 6.918E-01 3.876E+01 3.945E+01 A From Table 15 B From Table 16 I --

I Nuclear Common Revision 9 1 I Nu l a o m nR v s o

CALCULATION CONTINUATION SHEET SHEET 67 of 78 CALC. NO.: S-C-ZZ-MDC-1945

REFERENCE:

LCR S03-05 G. Patel/NUCORE, ORIGINATOR, DATE REV:

02/24/04 0

M Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/26/04 Table 18 0-1 Hr Isotopic Activity Released To Environment 0-1 hr Isotopic Integrated Total 0-1 hr Isotopic Integrated Total Isotope Activity Released To Integrated Isotope Activity Released To Integrated Environment (Ci)

Environment (Ci)

Containment ESF Activity Containment ESF Activity Leakage Leakage (Ci)

Leakage Leakage (Ci)

A B

A+B A

B A+B+C Co-58 6.884E-03 0.000E+00 6.884E-03 Te-131m 1.931E+00 0.000E+00 1.931E+00 Co-60 5.269E-03 0.000E+00 5.269E-03 Te-132 1.965E+01 0.000E+00 1.965E+01 Kr-85 5.960E+00 0.000E+00 5.960E+00 1-131 1.678E+02 1.412E+02 3.090E+02 Kr-85m 1.153E+02 0.OOOE+00 1.153E+02 1-132 1.666E+02 6.088E+01 2.274E+02 Kr-87 1.280E+02 0.000E+00 1.280E+02 1-133 3.269E+02 1.114E+02 4.383E+02 Kr-88 2.652E+02 0.000E+00 2.652E+02 I-134 1.529E+02 6.089E+01 2.138E+02 Rb-86 7.720E-02 0.000E+00 7.720E-02 1-135 2.849E+02 9.909E+01 3.839E+02 Sr-89 6.133E+00 0.OOOE+00 6.133E+00 Xe-133 1.076E+03 O.OOOE+00 1.076E+03 Sr-90 3.313E-01 0.000E+00 3.313E-01 Xe-135 2.454E+02 0.000E+00 2.454E+02 Sr-91 7.064E+00 0.000E+00 7.064E+00 Cs-134 1.771E+01 O.OOOE+00 1.771E+01 Sr-92 5.578E+00 0.000E+00 5.578E+00 Cs-136 5.373E+00 0.000E+00 5.373E+00 Y-90 3.497E-03 0.OOOE+00 3.497E-03 Cs-137 9.900E+00 0.000E+00 9.900E+00 Y-91 7.473E-02 0.OOOE+00 7.473E-02 Ba-139 5.034E+00 0.000E+00 5.034E+00 Y-92 6.128E-02 0.000E+00 6.128E-02 Ba-140 1.059E+01 0.000E+00 1.059E+01 Y-93 8.402E-02 0.000E+00 8.402E-02 La-140 1.058E-01 0.000E+00 1.058E-01 Zr-95 9.441 E-02 0.OOOE+00 9.441E-02 La-141 7.626E-02 0.000E+00 7.626E-02 Zr-97 9.252E-02 0.OOOE+00 9.252E-02 La-142 4.875E-02 O.OOOE+00 4.875E-02 Nb-95 8.920E-02 0.000E+00 8.920E-02 Ce-141 2.412E-01 0.000E+00 2.412E-01 Mo-99 1.283E+00 0.000E+00 1.283E+00 Ce-143 2.275E-01 0.000E+00 2.275E-01 Tc-99m 9.447E-01 0.000E+00 9.447E-01 Ce-144 1.455E-01 0.000E+00 1.455E-01 Ru-103 9.696E-01 0.000E+00 9.696E-01 Pr-143 9.193E-02 0.000E+00 9.193E-02 Ru-105 4.984E-01 0.000E+00 4.984E-01 Nd-147 4.108E-02 0.000E+00 4.108E-02 Ru-106 2.205E-01 0.000E+00 2.205E-01 Np-239 2.712E+00 0.000E+00 2.712E+00 Rh-105 4.245E-01 0.000E+00 4.245E-01 Pu-238 1.566E-04 0.000E+00 1.566E-04 Sb-127 1.178E+00 0.000E+00 1.178E+00 Pu-239 3.532E-05 O.OOOE+00 3.532E-05 Sb-129 3.309E+00 0.000E+00 3.309E+00 Pu-240 4.454E-05 O.OOOE+00 4.454E-05 Te-127 1.028E+00 0.000E+00 1.028E+00 Pu-241 7.500E-03 O.OOOE+00 7.500E-03 Te-127m 1.523E-01 0.000E+00 1.523E-01 Am-241 1.982E-06 0.000E+00 1.982E-06 Te-129 1.612E+00 0.000E+00 1.612E+00 Cm-242 7.582E-04 0.000E+00 7.582E-04 Te-129m 1.043E+00 0.000E+00 1.043E+00 Cm-244 4.438E-05 0.000E+00 4.438E-05 A From RADTRAD Output File SI5OCS7SCLO1.oO B From RADTRAD Output File S75ESFIGPM.oO l Nuclear Common Revision 9 l

I CALCULATION CONTINUATION SHEET SHEET 68 of 78 CALC. NO.: S-C-ZZ-MDC-1945

REFERENCE:

LCR S03-05 G. Patel/NUCORE, ORIGINATOR, DATE REV:

02/24/04 0

M Drucker/NUCORE, REVIEWERJVERIFIER, DATE 02/26/04 Table 19 0-2 Hr Isotopic Activity Released To Environment 0-2 hr Isotopic Integrated Total 0-2 hr Isotopic Integrated Total Isotope Activity Released To Integrated Isotope Activity Released To Integrated Environment (Ci)

Environment (Ci)

Containment ESF Activity Containment ESF Activity Leakage Leakage (Ci)

Leakage Leakage (Ci)

A B

A+B A

B A+B+C Co-58 3.812E-02 0.OOOE+0O 3.812E-02 Te-131m 1.048E+O1 0.OOOE+OO 1.048E+O01 Co-60 2.919E-02 O.OOOE+OO 2.919E-02 Te-132 1.080E+02 O.OOOE+OO 1.080E+02 Kr-85 4.14 1E+O I O.OOOE+O0 4.141E+O1 1-131 6.477E+02 3.524E+02 1.OOOE+03 Kr-85m 6.810E+02 0.OOOE+OO 6.81 0E+02 1-132 4.840E+02 3.518E+02 8.358E+02 Kr-87 5.190E+02 O.OOOE+OO 5.190E+02 1-133 1.225E+03 6.870E+02 1.912E+03 Kr-88 1.431E+03 0.OOOE+OO 1.431E+03 1-134 3.048E+02 3.264E+02 6.312E+02 Rb-86 2.734E-01 O.OOOE+OO 2.734E-01 1-135 9.939E+02 5.994E+02 1.593E+03 Sr-89 3.396E+O1 O.OOOE+OO 3.396E+O1 Xe-133 7.43 1E+03 O.OOOE+OO 7.43 lE+03 Sr-90 1.836E+OO 0.OOOE+OO 1.836E+OO Xe-135 1.572E+03 O.OOOE+OO 1.572E+03 Sr-91 3.663E+O1 0.OOOE+OO 3.663E+O1 Cs-134 6.280E+O1 O.OOOE+0O 6.280E+O1 Sr-92 2.471E+O1 O.OOOE+0O 2.471E+O1 Cs-136 1.902E+O1 O.OOOE+OO 1.902E+O01 Y-90 1.918E-02 O.OOOE+OO 1.918E-02 Cs-137 3.512E+O1 O.OOOE+OO 3.512E+01 Y-91 4.138E-01 O.OOOE+OO 4.138E-O01 Ba-139 1.839E+O1 O.OOOE+OO 1.839E+01 Y-92 2.854E-O01 0.OOOE+OO 2.854E-O1 Ba-140 5.856E+O01 O.OOOE+OO 5.856E+O1 Y-93 4.373E-O01 O.OOOE+OO 4.373E-O1 La-140 5.770E-O01 O.OOOE-+OO 5.770E-0O Zr-95 5.228E-O1 O.OOOE+OO 5.228E-O01 La-141 3.61 IE-Ol O.OOOE+OO 3.61 lE-Ol Zr-97 4.937E-O01 O.OOOE+OO 4.937E-OI La-142 1.853E-O01 O.OOOE+OO 1.853E-O1 Nb-95 4.938E-O01 O.OOOE+OO 4.938E-01 Ce-141 1.335E+0O O.OOOE+OO 1.335E+OO Mo-99 7.037E+OO O.OOOE+OO 7.037E+OO Ce-143 1.236E+OO 0.OOOE+OO 1.236E+OO Tc-99m 4.718E+0O O.OOOE+OO 4.718E+O0 Ce-144 8.061E-O01 0.OOOE+OO 8.061E-Ol Ru-103 5.368E+0o O.OOOE+O0 5.368E+OO Pr-143 5.083E-O01 O.OOOE+0O 5.083E-O01 Ru-105 2.402E+0O O.OOOE+OO 2.402E+OO Nd-147 2.271E-O1 O.OOOE+OO 2.271 E-O1 Ru-106 1.222E+OO O.OOOE+OO 1.222E+OO Np-239 1.486E+O1 O.OOOE+OO 1.486E+O01 Rh-105 2.310E+OO O.OOOE+OO 2.31 0E+OO Pu-238 8.676E-04 0.OOOE+OO 8.676E-04 Sb-127 6.481E+OO O.OOOE+OO0 6.481E+OO Pu-239 1.957E-04 O.OOOE+OO 1.957E-04 Sb-129 1.589E+O1 O.OOOE+OO 1.589E+O01 Pu-240 2.468E-04 O.OOOE+OO 2.468E-04 Te-127 5.324E+OO O.OOOE+OO 5.324E+OO Pu-241 4.155E-02 O.OOOE+OO 4.155E-02 Te-127m 8.434E-O01 O.OOOE+OO 8.434E-O1 Am-241 1.098E-05 O.OOOE+OO 1.098E-05 Te-129 5.502E+oo O.OOOE+OO 5.502E+OO Cm-242 4.200E-03 O.OOOE+OO 4.200E-03 Te-129m 5.773E+OO O.OOOE+OO 5.773E+OO Cm-244 2.459E-04 O.OOOE+OO 2.459E-04 A From RADTRAD Output File S150CS75CLOl.oO B From RADTRAD Output File S75ESFIGPM.oO I Nuclear Common Revision 9 1 Nu l a Co m nR v s o

CALCULATION CONTINUATION SHEET SHEET 69 of 78 CALC. NO.: S-C-ZZ-MDC-1945

REFERENCE:

LCR S03-05 G. Patel/NUCORE, ORIGINATOR, DATE REV:

02/24/04 0

M Drucker/NUCORE, REVIEWERtVERIFIER, DATE 02/26/04 Table 20 0-4 Hr Isotopic Activity Released To Environment 0-4 hr Isotopic Integrated Total 0-4 hr Isotopic Integrated Total Isotope Activity Released To Integrated Isotope Activity Released To Integrated Environment (Ci)

Environment (Ci)

Containment ESF Activity Containment ESF Activity Leakage Leakage (Ci)

Leakage Leakage (Ci)

A B

A+B A

B A+B Co-58 5.459E-02 0.000E+00 5.459E-02 Te-131m 1.484E+01 0.OOOE+00 1.484E+01 Co-60 4.181lE-02 0.000E+00 4.181E-02 Te-132 1.540E+02 0.000E+00 1.540E+02 Kr-85 1.340E+02 0.000E+00 1.340E+02 I-131 9.024E+02 7.725E+02 1.675E+03 Kr-85m 1.845E+03 0.000E+00 1.845E+03 1-132 5.955E+02 5.950E+02 1.190E+03 Kr-87 9.71 0E+02 0.000E+00 9.71 0E+02 1-133 1.682E+03 1.462E+03 3.144E+03 Kr-88 3.526E+03 0.000E+00 3.526E+03 1-134 3.325E+02 4.202E+02 7.526E+02 Rb-86 3.736E-01 0.000E+00 3.736E-01 I-135 1.321E+03 1.193E+03 2.513E+03 Sr-89 4.863E+01 0.000E+00 4.863E+01 Xe-133 2.388E+04 0.000E+00 2.388E+04 Sr-90 2.629E+00 0.000E+00 2.629E+00 Xe-135 4.650E+03 0.000E+00 4.650E+03 Sr-91 5.074E+01 0.0003E+00 5.074E+01 Cs-134 8.588E+01 0.000E+00 8.588E+01 Sr-92 3.185E+01 0.000E+00 3.185E+01 Cs-136 2.598E+01 0.000E+00 2.598E+01 Y-90 2.733E-02 0.000E+00 2.733E-02 Cs-137 4.802E+01 0.000E+00 4.802E+01 Y-91 5.926E-01 0.000E+00 5.926E-01 Ba-139 2.198E+01 0.OOOE+00 2.198E+01 Y-92 3.761E-01 0.000E+00 3.7611E-01 Ba-140 8.378E+01 0.000E+00 8.378E+01 Y-93 6.070E-01 0.000E+00 6.070E-01 La-140 8.197E-01 0.000E+00 8.197E-01 Zr-95 7.486E-01 0.000E+00 7.486E-01 La-141 4.794E-01 0.000E+00 4.794E-01 Zr-97 6.938E-01 0.000E+00 6.938E-01 La-142 2.247E-01 0.OOOE+00 2.247E-01 Nb-95 7.070E-01 0.OOOE+00 7.070E-01 Ce-141 1.912E+00 0.OOOE+00 1.912E+00 Mo-99 1.003E+01 0.000E+00 1.003E+01 Ce-143 1.753E+00 0.000E+00 1.753E+00 Tc-99m 6.419E+00 0.000E+00 6.419E+00 Ce-144 1.154E+00 0.000E+00 1.154E+00 Ru-103 7.686E+00 0.OOOE+00 7.686E+00 Pr-143 7.273E-01 0.000E+00 7.273E-01 Ru-105 3.213E+00 0.000E+00 3.213E+00 Nd-147 3.248E-01 0.000E+00 3.248E-01 Ru-106 1.750E+00 0.000E+00 1.750E+00 Np-239 2.115E+01 0.000E+00 2.115E+01 Rh-105 3.278E+00 0.000E+00 3.278E+00 Pu-238 1.243E-03 0.OOOE+00 1.243E-03 Sb-127 9.249E+00 0.OOOE+00 9.249E+00 Pu-239 2.802E-04 0.000E+00 2.802E-04 Sb-129 2.122E+01 0.OOOE+00 2.122E+01 Pu-240 3.534E-04 0.000E+00 3.534E-04 Te-127 7.372E+00 0.000E+00 7.372E+00 Pu-241 5.951E-02 0.000E+00 5.951E-02 Te-127m 1.208E+00 0.OOOE+00 1.208E+00 Am-241 1.572E-05 0.OOOE+00 1.572E-05 Te-129 6.427E+00 0.000E+00 6.427E+00 Cm-242 6.015E-03 0.OOOE+00 6.015E-03 Te-129m 8.264E+00 0.000E+00 8.264E+00 Cm-244 3.521E-04 0.000E+00 3.521E-04 A From RADTRAD Output File S150CS75CLOI.oO B From RADTRAD Output File S75ESFlGPM.oO I Nuclear Common Revision 9 1 I Nuclear Common Revision 9 I

I CALCULATION CONTINUATION SHEET SHEET 70 of 78 CALC. NO.: S-C-ZZ-MDC-1945

REFERENCE:

LCR S03-05 G. Patel/NUCORE, ORIGINATOR, DATE REV:

02/24/04 0

M Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/26/04 Table 21 0-8 Hr Isotopic Activity Released To Environment 0-8 hr Isotopic Integrated Total 0-8 hr Isotopic Integrated Total Isotope Activity Released To Integrated Isotope Activity Released To Integrated Environment (Ci)

Environment (Ci)

Containment ESF Activity Containment ESF Activity Leakage Leakage (Ci)

Leakage Leakage (Ci)

A B

A+B A

B A+B Co-58 6.768E-02 O.OOOE+00 6.768E-02 Te-131m 1.807E+Ol1 0.OOOE+00 1.807E+O1 Co-60 5.185E-02 0.OOOE+O0 5.185E-02 Te-132 1.896E+02 0.OOOE+0O 1.896E+02 Kr-85 3.190E+02 O.OOOE+0O 3.190E+02 I-131 1.120E+03 1.604E+03 2.724E+03 Kr-85m 3.326E+03 0.OOOE+00 3.326E+03 1-132 6.340E+02 8.009E+02 1.435E+03 Kr-87 1.174E+03 0.OOOE+00 1.174E+03 1-133 2.037E+03 2.866E+03 4.903E+03 Kr-88 5.601E+03 O.OOOE+00 5.601E+03 I-134 3.350E+02 4.434E+02 7.784E+02 Rb-86 4.531E-O1 0.OOOE+00 4.531E-OI I-135 1.524E+03 2.063E+03 3.587E+03 Sr-89 6.028E+01 0.OOOE+00 6.028E+01 Xe-133 5.624E+04 0.OOOE+00 5.624E+04 Sr-90 3.261 E+O0 O.OOOE+00 3.261 E+O0 Xe-135 9.561E+03 0.OOOE+00 9.561E+03 Sr-91 5.966E+01 0.OOOE+OO 5.966E+01 Cs-134 1.043E+02 0.OOOE+00 1.043E+02 Sr-92 3.443E+01 0.OOOE+00 3.443E+01 Cs-136 3.149E+01 0.OOOE+00 3.149E+01 Y-90 3.360E-02 0.OOOE+00 3.360E-02 Cs-137 5.83 1E+O1 O.OOOE+00 5.83 1E+01 Y-91 7.347E-O1 O.OOOE+00 7.347E-01 Ba-139 2.263E+01 O.OOOE+00 2.263E+01 Y-92 4.154E-O1 O.OOOE+0O 4.154E-01 Ba-140 1.037E+02 O.OOOE+O0 1.037E+02 Y-93 7.157E-01 0.OOOE+00 7.157E-01 La-140 1.003E+00 O.OOOE+00 1.003E+00 Zr-95 9.282E-0I 0.OOOE+00 9.282E-01 La-141 5.337E-0 I 0.OOOE+00 5.337E-01 Zr-97 8.335E-01 0.OOOE+O0 8.335E-01 La-142 2.329E-01 0.OOOE+0O 2.329E-01 Nb-95 8.763E-O1 0I.OOOE+00 8.763E-01 Ce-141 2.369E+0O 0.OOOE+00 2.369E+00 Mo-99 1.233E+01 0.OOOE+00 1.233E+01 Ce-143 2.138Ef+00 0.OOOE+00 2.138E+00 Tc-99m 7.361E+00 O.OOOE+00 7.361E+00 Ce-144 1.432E+00 0.OOOE+00 1.432E+00 Ru-103 9.527E+O0 O.OOOE+0O 9.527E+00 Pr-143 9.004E-01 O.OOOE+00 9.004E-01 Ru-105 3.610E+OO 0.OOOE+00 3.610E+0O Nd-147 4.019E-01 O.OOOE+00 4.019E-01 Ru-106 2.170E+OO O.OOOE+OO 2.170E+0O Np-239 2.597E+01 O.OOOE+00 2.597E+01 Rh-105 4.002E+OO O.OOOE+OO 4.002E+OO Pu-238 1.541E-03 0.OOOE+0O 1.541E-03 Sb-127 1.140E+01 0.OOOE+OO 1.140E+01 Pu-239 3.476E-04 O.OOOE+00.

3.476E-04 Sb-129 2.379E+O1 0.OOOE+OO 2.379E+O1 Pu-240 4.383E-04 O.OOOE+OOl 4.383E-04 Te-127 8.662E+O0 O.OOOE+OO 8.662E+00 Pu-241 7.381 E-02 0.OOOE+00 7.3811E-02 Te-127m 1.498E+OO 0.OOOE+OO 1.498E+O0 Am-241 1.950E-05 0.OOOE+OO 1.950E-05 Te-129 6.556E+0O 0.OOOE+OO 6.556E+OO Cm-242 7.459E-03 O.OOOE+O0 7.459E-03 Te-129m 1.024E+O1 O.OOOE+0O 1.024E+01 Cm-244 4.367E-04 0.OOOE+O0 4.367E-04 A From RADTRAD Output File S150CS75CLOl.oO B From RADTRAD Output File S75ESFlGPM.oO j ~

cl ar Co mo R viio_

I Nuclear Common Revision 9 1

CALCULATION CONTINUATION SHEET SHEET 71 of 78 CALC. NO.: S-C-ZZ-MDC-1945

REFERENCE:

LCR S03-05 RV G. Patel/NUCORE, ORIGINATOR, DATE REV:

02124/04 0

M Drucker/NUCORE, REVIEWERJVERIFIER, DATE 02t26/04 Table 22 0-24 Hr Isotopic Activity Released To Environment 0-24 hr Isotopic Integrated Total 0-24 hr Isotopic Integrated Total Isotope Activity Released To Integrated Isotope Activity Released To Integrated Environment (Ci)

Environment (Ci)

Containment ESF Activity Containment ESF Activity Leakage Leakage (Ci)

Leakage Leakage (Ci)

A B

A+B A

B A+B Co-58 9.387E-02 0.000E+00 9.387E-02 Te-131m 2.332E+01 0.000E+00 2.332E+01 Co-60 7.202E-02 0.000E+00 7.202E-02 Te-132 2.554E+02 0.000E+00 2.554E+02 Kr-85 1.059E+03 0.000E+00 1.059E+03 I-131 1.599E+03 4.803E+03 6.401E+03 Kr-85m 4.823E+03 0.OOOE+00 4.823E+03 1-132 6.448E+02 8.839E+02 1.529E+03 Kr-87 1.199E+03 O.OOOE+00 1.199E+03 1-133 2.630E+03 6.857E+03 9.487E+03 Kr-88 6.762E+03 0.000E+00 6.762E+03 I-134 3.35 1E+02 4.444E+02 7.795E+02 Rb-86 6.103E-01 0.000E+00 6.103E-01 1-135 1.713E+03 3.360E+03 5.074E+03 Sr-89 8.356E+01 0.000E+00 8.356E+01 Xe-133 1.783E+05 0.000E+00 1.783E+05 Sr-90 4.529E+00 0.000E+00 4.529E+00 Xe-135 1.890E+04 0.000E+00 1.890E+04 Sr-91 6.925E+01 0.000E+00 6.925E+01 Cs-134 1.412E+02 0.000E+00 1.412E+02 Sr-92 3.529E+01 0.000E+00 3.529E+01 Cs-136 4.232E+01 0.000E+00 4.232E+01 Y-90 4.499E-02 0.000E+00 4.499E-02 Cs-137 7.896E+01 0.OOOE+00 7.896E+01 Y-91 1.019E+00 0.000E+00 1.019E+00 Ba-139 2.269E+01 0.OOOE+00 2.269E+01 Y-92 4.337E-01 0.000E+00 4.337E-01 Ba-140 1.429E+02 0.000E+00 1.429E+02 Y-93 8.366E-01 0.000E+00 8.366E-01 La-140 1.316E+00 0.000E+00 1.316E+00 Zr-95 1.287E+00 0.000E+00 1.287E+00 La-141 5.623E-01 0.000E+00 5.623E-01 Zr-97 1.028E+00 0.000E+00 1.028E+00 La-142 2.339E-01 0.000E+00 2.339E-01 Nb-95 1.214E+00 0.000E+00 1.214E+00 Ce-141 3.280E+00 0.000E+00 3.280E+00 Mo-99 1.653E+01 0.000E+00 1.653E+01 Ce-143 2.776E+00 0.000E+00 2.776E+00 Tc-99m 8.103E+00 0.OOOE+00 8.103E+00 Ce-144 1.988E+00 0.000E+00 1.988E+00 Ru-103 1.320E+01 0.000E+00 1.320E+01 Pr-143 1.241E+00 0.000E+00 1.241E+00 Ru-105 3.847E+00 0.000E+00 3.847E+00 Nd-147 5.531E-01 0.000E+00 5.53 lE-01 Ru-106 3.013E+00 0.000E+00 3.013E+00 Np-239 3.462E+01 0.000E+00 3.462E+01 Rh-105 5.217E+00 O.OOOE+00 5.217E+00 Pu-238 2.141E-03 0.000E+00 2.141E-03 Sb-127 1.543E+01 0.000E+00 1.543E+01 Pu-239 4.828E-04 0.OOOE+00 4.828E-04 Sb-129 2.529E+01 0.000E+00 2.529E+01 Pu-240 6.088E-04 0.OOOE+00 6.088E-04 Te-127 1.004E+01 0.000E+00 1.004E+01 Pu-241 1.025E-01 0.000E+00 1.025E-01 Te-127m 2.078E+00 0.000E+00 2.078E+00 Am-241 2.709E-05 0.OOOE+00 2.709E-05 Te-129 6.564E+00 0.000E+00 6.564E+00 Cm-242 1.035E-02 0.000E+00 1.035E-02 Te-129m 1.418E+01 0.000E+00 1.418E1+01 Cm-244 6.066E-04 0.000E+00 6.066E-04 A From RADTRAD OutputFile S150CS75CLO1.oO B From RADTRAD Output File S75ESFlGPM.oO I Nuclear Common Revision 9 1 Nuclear Common Revision 9

I

,TlCALCULATION CONTINUATION SHEET l SHEET 72 of 78 CALC. NO.: S-C-ZZ-MDC-1945

REFERENCE:

LCR S03-05 G. Patel/NUCORE, ORIGINATOR, DATE IREV:

02/24/04 0

l l

M Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02126/04 11.0 FIGURES:

Figure 1: Containment Leakage RADTRAD Nodalization INuclear Common Revision 9 I Nuclear Common Revision 9 I

I CALCULATION CONTINUATION SHEET ISHEET 73 of 78 CALC. NO.: S-C-ZZ-MDC-1945

REFERENCE:

LCR S03-05 l

G. Patel/NUCORE, ORIGINATOR, DATE I REV:

02/24/04 0

l l

M Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02t26/04 Figure 2: Containment ESF Leakage RADTRAD Nodalization l Nuclear Common Revision 9 I Nuclear Common Revision 9 I

I CALCULATION CONTINUATION SHEET SHEET 74 of 78 CALC. NO.: S-C-ZZ-MDC-1945

REFERENCE:

LCR S03-05 G. Patel/NUCORE, ORIGINATOR, DATE REV:

02/24/04 0

M Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/26/04 Figure 3 - Salem Control Room RADTRAD Nodalization INuclear Common Revision 9 Nuclear Common Revision 9 I

I CALCULATION CONTINUATION SHEET ISHEET 75 of 78 CALC. NO.: S-C-ZZ-MDC-1945

REFERENCE:

LCR S03-05 G. Patel/NUCORE, ORIGINATOR, DATE REV:

02-2404 0l M Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/26104 J!

97'-6" S480 Unit 2 CR Intake EL 129'-5-1/4" Unit 1 CR Intake EL 130'-3-1/2" IS326 Figure 4: Locations Of RWST With Respect Units 1 & 2 CR Air Intakes & Plant Vents I Nuclear Common Revision 9 1 Nuclear Common Revision 9 I

I CALCULATION CONTINUATION SHEET ISHEET 76 of 78 CALC. NO.: S-C-ZZ-MDC-1945

REFERENCE:

LCR S03-05 G. Patel/NUCORE, ORIGINATOR, DATE REV:

02/24/04 0

M Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/26/04 Plant North eometry Distance To Release Point Direction CR Release Receptor Height To Wake Intake Point feet meter feet meter Source Area Height degree In Meter Unit 1 Plant Vent 137.42 41.9 192.00 58.54 189.08 2429.54 12.35 Center of CR l Nuclear Common Revision 9 l Nuclear Common Revision 9

CALCULATION CONTINUATION SHEET SHEET 77 of 78 CALC. NO.: S-C-ZZ-MDC-1945

REFERENCE:

LCR S03-05 G. Patel/NUCORE, ORIGINATOR, DATE REV:

02/24/04 0

l M Drucker/NUCORE, REVIEWERNVERIFIER, DATE 02/26/04 Conference Room

.1h OSC

'1 Dose Point In CR 7-ff21-0" 37'-7-1/2" 1 0'4-1/2" 8'-6" Floor Elevation = 122'-0" Top of Charcoal Filter Bed = 136'-7" Bottom Charc Height Charco oal Filter Bed = 1329-4" Y

' Z oal = 4'-3" I l 296" Figure 6: MicroShield Geometry Model - Post-LOCA Charcoal Filter Shine Dose Nuclear Common Revision 9 I Nuclear Common Revision 9

I CALCULATION CONTINUATION SHEET SHEET 78 of 78 CALC. NO.: S-C-ZZ-MDC-1945

REFERENCE:

LCRS03-05 l

G. Patel/NUCORE, ORIGINATOR, DATE REV:

02t24/04 0

M Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02t26/04 12.0 AFFECTED DOCUMENTS:

Upon approval of Licensing Change Request LCR S03-05, the following documents will be revised:

UFSAR Section 15.4.1, Major Reactor Coolant System Pipe Ruptures (Loss-Of-Coolant Accident)

UFSAR Table 15.4-SA, Physical Data for Isotopes UFSAR Table 15.4-SB, Loss-Of-Coolant Analysis Parameter and Assumptions UFSAR Table 15.4-SC, Loss-Of-Coolant Accident Dose Consequences UFSAR Table 15.4-SD, Salem Control Room Dispersion Factors for LOCA Analysis UFSAR Table 15.4-SE, Salem Control Room Doses From a Loss-Of-Coolant Accident The comparison of doses in Sections 7.1 & 7.2 indicates that the containment spray interruption increases the CR dose less than 1%. The containment leakage model is used in other calculations including S-C-ZZ-MDC-1946 (TSC Habitability), S-C-ZZ-MDC-1947 (Vital Access Area Mission Dose), and S-C-ZZ-MDC-2005 (Hope Creek CR Habitability). The increase in the CR dose due to the containment spray interruption is < 1%,

therefore, its impact on TSC and Hope Creek CR habitability is considered negligible. The EQ doses in S-C-ZZ-MDC-2008 uses the post-LOCA sump water activity, which is not impacted by the containment spray interruption. The Vital Access Area Mission airborne dose uses the LPZ dose model. The LPZ dose increases by 2.5%.Therefore, the impact of containment spray interruption on the Vital Access Area Mission is not expected to increase more than 2.5%, which is considered negligible.

13.0 ATTACHMENTS:

2 Diskettes with the various electronic files.

Calculation No: S-C-ZZ-MDC-1945, Rev OIR1 Comment Resolution Form 2 - Mark Drucker Certification for Design Verification Form-1 RCPD Form-l I Nuclear Common Revision 9 1 Nuclear Common Revision 9

S-C-ZZ-MDC-1945, Rev 0 3.0 2 Diskettes With Various Electronic Files