ML042740511

Eab, LPZ, & CR Doses - Rod Ejection Accident (Rea) - AST

ML042740511
Person / Time
Site:	Salem
Issue date:	02/09/2004
From:	Gita Patel NUCORE
To:	Office of Nuclear Reactor Regulation
References
LR-N04-0413 S-C-ZZ-MDC-1948, Rev OIR2
Download: ML042740511 (35)

Text

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

CALC NO.: S-C-ZZ-MDC-1948 CALCULATION COVER SHEET Page 1 of 33 REVISION: 01R2 CALC. TITLE:

EAB, LPZ, & CR Doses - Rod Ejection Accident (REA) - AST

1. SHTS (CALC): _ 3 X ATT #SHTS: I1/1 l #IDV150.59 SHTS: l2/3.

l # TOTAL SHTS: l 4 -48 CHECK ONE:

1 3 k El FINAL 0 INTERIM (Proposed Plant Change)

El FINAL (Future Confirmation Req'd)

[1 VOID SALEM OR HOPE CREEK:

D Q-LIST

[9 IMPORTANT TO SAFETY El NON-SAFETY RELATED HOPE CREEK ONLY:

nQ flQs nQsh OF EOR El STATION PROCEDURES IMPACTED, IF SO CONTACT RELIABILITY ENGINEER El CDs INCORPORATED (IF ANY):

DESCRIPTION OF CALCULATION REVISION (IF APPL.):

N/A PURPOSE:

The purpose of this calculation is to determine the Exclusion Area Boundary (EAB), Low Population Zone (LPZ), and Control Room (CR) doses due to a Rod Ejection Accident (REA) using the Alternative Source Term (AST) methodology and total effective dose equivalent (TEDE) dose criteria. The dose consequences are representative of an accident occurring in either unit.

CONCLUSIONS:

The Rod Ejection Accident results presented in Section 7.0 indicate that the EAB, LPZ, and CR doses due to a REA are within their allowable limits.

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

CALC NO.: S-C-ZZ-MDC-1948 CALCULATION COVER SHEET Page 1 of 33 REVISION: OlR2 CALC. TITLE:

EAB, LPZ, & CR Doses - Rod Ejection Accident (REA) - AST

1. SHTS (CALC):

l 33

1. ATT I # SHTS:

111

1. IDVI50.59 SHTS:

l 213

1. TOTAL SHTS:

39 CHECK ONE:

n FINAL 0 INTERIM (Proposed Plant Change)

D FINAL (Future Confirmation Req'd) 5 VOID SALEM OR HOPE CREEK:

5 Q - LIST N IMPORTANT TO SAFETY 5 NON-SAFETY RELATED HOPE CREEK ONLY:

51Q nQs flQsh OF OR 5 STATION PROCEDURES IMPACTED, IF SO CONTACT RELIABILITY ENGINEER 5

CDs INCORPORATED (IF ANY):

DESCRIPTION OF CALCULATION REVISION IF APPL.):

N/A PURPOSE:

The purpose of this calculation is to determine the Exclusion Area Boundary (EAB), Low Population Zone (LPZ), and Control Room (CR) doses due to a Rod Ejection Accident (REA) using the Alternative Source Term (AST) methodology and total effective dose equivalent (TEDE) dose criteria. The dose consequences are representative of an accident occurring in either unit.

CONCLUSIONS:

The Rod Ejection Accident results presented in Section 7.0 indicate that the EAB, LPZ, and CR doses due to a REA are within their allowable limits.

Printed Name / Signature Date ORIGINATOR/COMPANY NAME:

Gopal J. Patel/NUCORE 02/09/04 REVIEWER/COMPANY NAME:

N/A N/A VERIFIER/COMPANY NAME:

Mark Drucker/NUCORE 02/09/04 PSEG SUPERVISOR APPROVAL:

Paul Lindsay/PSEG I Nuclear Common Revision 91

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Mark Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/09/04 REVISION HISTORY Revision Revision Description OIRO Initial Issue.

OIR l Editorial changes incorporated.

OIR2 Design validation comments incorporated.

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Mark Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/09/04 TABLE OF CONTENTS Section Sheet No.

Cover Sheet 1

Revision History 2

Page Revision Index 3

Table of Contents 4

1.0 Purpose 5

2.0 Background

5 3.0 Analytical Approach 5

4.0 Assumptions 8

5.0 Design Inputs 14 6.0 Calculations 19 7.0 Results Summary 23 8.0 Conclusions 23 9.0 References 24 10.0 Tables 26 11.0 Figures 30 12.0 Affected Documents 33 13.0 Attachments 33 I Nuclear Common Revision 9 l

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Mark Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/09/04

1.0 PURPOSE

The purpose of this calculation is to determine the Exclusion Area Boundary (EAB), Low Population Zone (LPZ), and Control Room (CR) doses due to a Rod Ejection Accident (REA) using the Alternative Source Term (AST) methodology and total effective dose equivalent (TEDE) dose criteria. The dose consequences are representative of an accident occurring in either unit.

2.0 BACKGROUND

The consequences of a REA 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 100.11. Also, the 5 rem TEDE control room dose acceptance criterion specified in 10 CFR 50.67 (Ref. 9.3) 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 REA is analyzed using plant specific design and licensing bases inputs which are compatible to the TEDE dose criteria. The REA analysis is performed using the guidance in Regulatory Guide 1.183 and its Appendix H (Ref. 9.1).

3.0 ANALYTICAL APPROACH:

This analysis uses Version 3.02 of the RADTRAD computer code (Ref.9.2) to calculate the potential radiological consequences of the REA. The RADTRAD code is documented in NUREG/CR-6604 (Ref. 9.2).

The RADTRAD code is maintained as Software ID Number A-0-ZZ-MCS-0225 (Ref. 9.4).

The calculation assumes that the CR air intake monitors preferentially select the less contaminated air intake when only one CREACS train is available. There are two CREACS trains that each provide emergency filtration and air conditioning services to the combined CR for the Salem 1 & 2 plants (Ref. 9.16). Each CREACS train is safety related and required to be operable by a Technical Specification (Refs. 9.6.12 and 9.6.13). Each CREACS train takes outside air supplied through two independent intakes, has safety related fans and associated radiation monitors (Ref. 9.16), which make the CREACS air supply system redundant. The redundant air intake monitors preferentially select the less contaminated air intake during an accident condition when the radiation level at the normal intake exceeds the setpoint value (see Section 6.5 for CR intake monitor setpoint evaluation). Based on the Salem plant-specific CREACS design and performance, the post-accident CREACS response is credited in the analysis with the CR air intake monitor's ability to preferentially select the less contaminated air intake.

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Mark Drucker/NUCORE, REVIEWERNERIFIER, DATE 02/09/04 Per Section 4.0, this calculation addresses two release cases. In the first, 100% of the activity released from the fuel is assumed to be released instantaneously and homogeneously throughout the containment atmosphere and available for release to the environment. In the second, 100% of the activity released from the fuel is assumed to be completely dissolved in the primary coolant (PC) and available for release to the environment through the secondary system.

3.1 Containment Leakage Release:

The following activity is assumed to be instantaneously and homogeneously distributed in the containment following a control rod ejection accident (Ref. 9.1, Appendix H, Section 1):

10% of the core iodine and 10% of the core noble gases in the fuel gap of 10% failed fuel, 25% of the core iodine and 100% of the core noble gases in the 0.25% melted fuel, and 100% of the iodine and noble gases initially present (i.e., pre-REA) in the reactor coolant system (RCS)

During the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> the containment is assumed to leak at its maximum technical specification leak rate of 0.10 volume percent per day and at 50% of this leak rate for the remaining duration of the accident (Ref. 9.1, Appendix H, Section 6.2, and Ref. 9.6.5). Although allowed by Regulatory Guide 1.183 (Ref. 9.1, Appendix H, Section 6.1), no credit is taken for a reduction in the amount of radioactive material available for leakage from the containment due to natural deposition and containment spray.

The post-REA activity released into the containment from the fuel and RCS is calculated in Tables 1 through 3.

The iodine and noble gas core inventory at a core power level of 3600 MWt is obtained from Reference 9.11 and listed in Table 1. The design basis core inventory is established in Table 1 based on scaling the 3600 MWt core inventory to an inventory based on 105% of the rated thermal power level of 3,459 MWt (Ref. 9.6.1). The pre-REA RCS activity available for release to the containment is calculated in Table 2 using the primary coolant activity concentrations based on 1% fuel defects (Ref. 9.11, Table 4) and RCS mass (Section 6.1). The post-REA iodine and noble gas activity released from the 10% failed fuel and 0.25% melted fuel are calculated in Table 3 with appropriate release fractions (Ref. 9.1, Appendix H, Section 1) applied to the Table 1 design basis core inventory. The activities released from the fuel and RCS into the containment are summed in Table 3, and input into RADTRAD Nuclide Inventory File (NIF) SREACNT def.txt.

The post-REA containment leakage release activity transport and CR response models are shown in Figures 1 &

3, which are used to determine the post-REA radiological consequences using the Salem plant-specific as-built design inputs and assumptions listed in Sections 4 & 5. The post-REA containment leakage release CR response model takes credit for CR operator action to manually initiate the CREACS operation 30 minutes after a REA because the activity concentration from this release path does not exceed the CR monitor setpoint (Section 6.5).

It is conservatively assumed that the CR HVAC draws outside air from the more contaminated CR air intake.

The EAB, LPZ, and CR doses are summarized in Section 7.0.

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Mark Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/09/04 3.2 Secondary System Releases To maximize the calculated secondary system release doses it is assumed that offsite power is lost so that the main steam condensers are not available. Due to the REA, the reactor is shutdown and the plant begins discharging secondary coolant via the steam generator (SG) relief valves. Steam release from the SGs continue for 110 seconds with the total steam mass release of 5.12E+05 lbs (Design Input 5.4.9). During the 110 seconds release period reactor coolant is assumed to leak into the SGs at the maximum technical specification rate of 1 gpm (Refs. 9.6.3 & 9.6.4). In the secondary system release case, 100% of the activity released from the fuel is assumed to be completely dissolved in the primary coolant (PC) and available for release to the SGs via primary-to-secondary (P-T-S) leakage (Ref. 9.1, Appendix H, Section 3). The reactor coolant noble gases that enter the SGs via P-T-S leakage are released directly to the environment without reduction and mitigation (Ref.

9.1, Appendix H, Section 7.3). It is conservatively assumed that the SG tubes are uncovered (i.e., not submerged in the SGs liquid), and consequently iodine introduced into the SGs via the P-T-S leakage is assumed to be directly released to the environment in proportion to the steam release rate, with no credit taken for iodine partitioning in the SG liquid. Although the affected unit CR intake preferentially aligns with the less contaminated air intake in 1 minute after the REA (see Section 6.5 for the CR monitor response), the 0-2 hour Unit 1 CR air intake X/Q for MSSV Set 1 release is conservatively used for the P-T-S leakage iodine release for entire duration of 110 seconds without crediting the CREACS ability to align with less contaminated air intake.

The post-REA iodine and noble gas activities released from the 10% failed fuel and 0.25% melted fuel, which are completely dissolved in the PC and available for release to SGs, are calculated in Table 4 with appropriate release fractions (Ref. 9.1, Appendix H, Section 1), and the RADTRAD NIF SREARCSdef.txt is developed using the total activity released to PC. The post-REA secondary system release activity transport and CR response models are shown in Figures 2 & 3, which are used to determine the post-REA radiological consequences resulting from the P-T-S leakage. The post-REA secondary system release CR response model utilizes the Salem CR air intake monitors to preferentially select the less contaminated air intake (Ref. 9.10, Section 3.7) because the activity concentration from this release path instantly exceeds the CR monitor setpoint (Section 6.5). The assumed delay time of 1 minute for the CR pressurization mode to be fully operational is conservatively assumed.

The EAB, LPZ, and CR doses from the containment leakage and secondary system releases are combined in Section 7.0 as indicated by Reference 9.13, Section 3.6, which states that the actual doses from a REA would be a composite of the two pathways.

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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 PWR rod ejection accident. The following sections address the applicability of these modeling assumptions to an SNGS Units I and 2 REA analysis.

These assumptions are incorporated as design inputs in Sections 5.3 through 5.6.

The radioactivity 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 exclusion area boundary (EAB) and the low population zone (LPZ) TEDE values (Ref 9.1, Section 4.2.2).

4.1 Source Term 4.1.1 Fuel Damage Source Term Per Reference 9.12, fuel damage is postulated for the REA. Ten percent (10%) of the core is assumed to experience clad damage and one-fourth of one percent (0.25%) of the core is assumed to experience fuel melt as a result of the REA.

Consistent with RG 1.183 (Ref. 9.1, Appendix H, Section 1), the release from the breached fuel to the containment and available for containment release, or to the primary coolant and available for release to the secondary system, is based on the estimate of the number of fuel rods breached (i.e., 10%) and the assumption that 10% of the core inventory of the noble gases and iodines is in the fuel gap.

Consistent with RG 1.183 (Ref. 9.1, Appendix H, Section 1), the release attributed to fuel melting is based on the fraction of the fuel (i.e., 0.25%) that reaches or exceed the initiation temperature for fuel melting and the assumption that 100% of the noble gases and 25% of the iodines contained in that fraction are available for release from containment.

Consistent with RG 1.183 (Ref. 9.1, Appendix H, Section 1), for the secondary system release pathway, 100%

of the noble gases and 50% of the iodines in the fraction of the fuel (i.e., 0.25%) that reaches or exceed the initiation temperature for fuel melting are released to the reactor coolant.

The above source term requirements are incorporated in the Design Inputs 5.3.1 through 5.3.9.

4.1.2 No Fuel Damage Source Term Consistent with RG 1.183 (Ref. 9.1, Appendix H, Section 2), since fuel damage is postulated for the REA a radiological analysis for the REA is required.

4.1.3 Release Cases to be Considered Consistent with RG 1.183 (Ref. 9.1, Appendix H, Section 3), two release cases are considered in this REA analysis. In the first, 100% of the activity released from the fuel is assumed to be released instantaneously and homogeneously through the containment atmosphere. In the second, 100% of the activity released from the fuel Nuclear Common Revision 9

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Mark Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/09/04 is assumed to be completely dissolved in the primary coolant and available for release to the secondary system (see Figures 1 & 2).

4.1.4 Containment Airborne Iodine Form Consistent with RG 1.183 (Ref. 9.1, Appendix H, Section 4), the chemical form of radioiodine released to the containment atmosphere is assumed to be 95% cesium iodide (CsI) (i.e., particulate), 4.85% elemental iodine, and 0.15% organic iodide (Design Input 5.4.4).

4.1.5 Steam Generator Release Iodine Form Consistent with RG 1.1 83 (Ref. 9.1, Appendix H, Section 5), iodine releases from the steam generators to the environment are assumed to be 97% elemental and 3% organic (Design Input 5.4.1 1).

4.2 Transport 4.2.1 Transport From Containment 4.2.1.1 Reduction in Radioactive Material RG 1.183 (Ref. 9.1, Appendix H, Section 6.1) allows for a reduction in the amount of radioactive material available for leakage from the containment that is due to natural deposition, containment sprays, recirculating filter systems, dual containments, or other engineered safety features to be taken into account. Conservatively, in this analysis no reduction in the post-REA airborne activity is credited.

4.2.1.2 Containment Leakage Consistent with RG 1.1 83 (Ref. 9. 1, Appendix H, Section 6.2) the containment is assumed to leak at the leak rate of 0.1 V%/day incorporated in the technical specifications at peak accident pressure for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and at 50% of this leak rate for the remaining duration of the accident. Peak accident pressure is the maximum pressure defined in the technical specifications for containment leak testing (Design Input 5.4.2).

4.2.2 Transport From Secondary System 4.2.2.1 Primary-to-Secondary Leak Rates Consistent with RG 1.183 (Ref. 9.1, Appendix H, Section 7.1) a leak rate equivalent to the primary-to-secondary leak rate limiting condition for operation specified in the technical specifications is assumed to exist until shutdown cooling is in operation and releases from the steam generators have been terminated (Design Inputs 5.4.9 & 5.4.10). Releases from the steam generators terminate in 1 10 seconds, once the full secondary side mass in all four steam generators has discharged through the main steam safety valves (Ref. 9.7, page 1 0).

4.2.2.2 Primary-to-Secondary Leak Primary Coolant Density Consistent with RG 1.183 (Ref. 9.1, Appendix H, Section 7.2) the primary coolant density used in converting the volumetric primary-to-secondary leak rates (e.g., gpm) to mass leak rates (e.g., lbmlhr) is assumed to be 1.0 gm/cc (62.4 lbm/ft3).

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4.2.2.3 Noble Gas Transport Model (Releases to the Environment)

Consistent with RG 1.183 (Ref. 9.1, Appendix H, Section 7.3) all noble gas radionuclides released from the primary system (via the primary-to-secondary leaks) are assumed to be released to the environment without reduction or mitigation.

4.2.2.4 Iodine and Particulate Transport Model (Releases to the Environment)

RG 1.183 (Ref. 9.1, Appendix H, Section 7.4) states that the transport model described in assumptions 5.5 and 5.6 of Appendix E should be utilized for iodine and particulate. In this analysis, all iodine leaked into the steam generators is diluted within the steam generator liquid mass, and released to the environment at the steam generator steaming rate assuming an iodine partition coefficient of one. Since the post-REA release is less than 2 minutes in duration, and since iodine partitioning is not credited in the analysis, use of the recommended transport model is not appropriate for the steam mass release from the intact SGs.

4.3 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 outer boundary of the low population zone (LPZ). The following sections address the applicability of this guidance to the SNGS Units 1 and 2 REA analysis. These assumptions are incorporated as design inputs in Section 5.6.1 through 5.6.6.

43.1 Modeling of Parent and Daughter Isotopes Consistent with RG 1.183 (Ref. 9.1, Section 4.1.1), this dose calculation determines the TEDE. 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 Sections 5.3.2 & 5.3.3.

4.3.2 CEDE (Inhalation) Dose Conversion Factors 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. 9.20).

4.3.3 Offsite Breathing Rates 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 104 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 Nuclear Common Revision 9

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Mark DruckerfNUCORE, REVIEWER/VERIFIER, DATE 02/09/04 assumed to be 2.3 x 104 cubic meters per second. These offsite breathing rates are listed in Sections 5.6.3 and 5.6.4.

4.3.4 DDE (Immersion) Dose Conversion Factors 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. 9.21).

4.3.5 Exclusion Area Boundary Dose Acceptance Criteria 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 CFR50.67 (Ref. 9.3). For the REA 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 5.6.4.

EAB Dose Acceptance Criterion:

6.3 Rem TEDE Per RG 1.183 Table 6, the REA event release duration is 30 days for the containment pathway, and until cold shutdown is established for the secondary pathway.

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 progression of the accident to capture the peak dose interval between the start of the event and the end of radioactivity release.

4.3.6 Low Population Zone Outer Boundary Dose Acceptance Criteria 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. 9.3). For the REA 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 5.6.5.

LPZ Dose Acceptance Criterion:

6.3 Rem TEDE 4.3.7 Effluent Plume Depletion 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.

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Mark Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/09/04 4.4 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 REA analysis. These assumptions are incorporated as design inputs in Sections 5.5.1 through 5.5.15.

4.4.1 Control Room Operator Dose Contributors 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].

• Note: The external airborne cloud shine dose, containment shine dose, and the CR filter shine dose due to a REA are insignificant compared to those due to a LOCA (see the core release fractions for LOCA and REA design basis accidents in Table 1 and Appendix H Section 1 of Reference 9. 1, respectively). Therefore, these direct dose contributions are considered to be insignificant and are not evaluated for a REA.

4.4.2 Control Room Source Term 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.

4.4.3 Control Room Transport Consistent with RG 1.183 (Ref. 9.1, Section 4.2.3), the model used to transport radioactive material into and through the control room is structured to provide suitably conservative estimates of the exposure to control room personnel. The shielding models are not developed to determine radiation dose rates from external sources because their dose contributions are insignificant compared to those from a LOCA (Section 4.4.1).

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Mark Drucker/NUCORE, REVIEWERNVERIFIER, DATE 02/09/04 4.4.4 Control Room Response 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. Control room isolation.is actuated by ESF signals or radiation monitors (RMs). In some cases, the ESF signal is effective only for selected accidents, placing reliance on the RMs for the remaining accidents (see Section 6.5 for CR intake monitor setpoint evaluation).

Several aspects of RMs can delay the control room isolation, including the delay for activity to build up to concentrations equivalent to the alarm setpoint and the effects of different radionuclide accident isotopic mixes on monitor response.

4.4.5 Control Room Operator Use of Dose Mitigating Devices 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) or prophylactic drugs (i.e., potassium iodide [KI] pills).

4.4.6 Control Room Occupancy Factors and Breathing Rates Consistent with RG 1.183 (Ref. 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 104 cubic meters per second. These assumptions are incorporated as design inputs in Sections 5.5.8 and 5.5.9.

4.4.7 Control Room Dose Conversion Factors 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 Regulatory Position 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, DDEfinit,, 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 = (DDEao x V0338) / 1173 4.4.8 Control Room Operator Dose Acceptance Criteria Consistent with RG 1.183 (Ref. 9.1, Section 4.4), for the CRA accident the postulated CR doses should not exceed the 5 Rem TEDE criterion established in 10 CFR 50.67 (Ref 9.3). This assumption is incorporated as a design input in Section 5.5.13:

CR Dose Acceptance Criterion:

5 Rem TEDE Nuclear Common Revision 9 l Nuclear Common Revision 9 l

I CALCULATION CONTINUATION SHEET ISHEET 14 of 33 CALC. NO.: S-C-ZZ-MDC-1948

REFERENCE:

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

02/09/04 0

Mark Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/09/04 5.0 DESIGN INPUTS:

5.1 General Considerations 5.1.1 Applicability of Prior Licensing Basis The implementation of an Alternative Source Term is a significant change to the design basis of the facility and to the assumptions and design inputs used in the analyses. The characteristics of the ASTs 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 plant specific design inputs and assumptions used in the current TID-1 4844 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 ASTs and the TEDE criteria.

5.1.2 Credit for Engineered Safety Features Credit is taken only for 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.

5.1.3 Assignment of Numeric Input Values The numeric values that are chosen as inputs to the analyses required by 10 CFR 50.67 (Ref. 9.3) are compatible to AST and TEDE dose criteria and selected with the objective of producing conservative radiological consequences. For conservatism, the limiting values of reactor coolant iodine concentrations listed in the SNGS Technical Specification are used in the analysis.

5.1.4 Meteorology Considerations The control room atmospheric dispersion factors (X/Qs) are developed in Reference 9.5 using the NRC sponsored ARCON96 computer code. The offsite X/Qs were accepted by the staff in previous licensing proceedings.

5.2 Accident-Specific Design Inputs/Assumptions The design inputs/assumptions utilized in the EAB, LPZ, and CR habitability analyses are listed in the following sections. The design inputs are compatible with the AST and TEDE dose criteria and assumptions are consistent with those identified in Appendix E of RG 1.183 (Ref. 9.1). The design inputs and assumptions in the following sections represent the as-built design of the plant.

INuclear Common Revision 9 l I Nuclear Common Revision 9 I

CALCULATION CONTINUATION SHEET SHEET 15 of 33 CALC. NO.: S-C-ZZ-MDC-1948

REFERENCE:

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

02/09/04 0

Mark Drucker/NUCORE, REVIEWERIVERIFIER, DATE 02/09/04 Design Input Parameter Value Assigned Reference 5.3 Rod Ejection Accident Parameters 5.3 Source Terms 5.3.1 Licensed power level 3,459 MWt 9.6.1 5.3.2 Core isotopic activity (at 3,600 MWt) 9.11, Table 2 Isotope Activity (Ci)

Isotope Activity (Ci)

Isotope Activity (Ci)

I-131 9.9E+07 Kr-85M 2.6E+07 Xe-133 2.OE+08 I-132 1.4E+08 Kr-85 1.lE+06 Xe-135M 4.OE+07 1-133 2.OE+08 Kr-87 4.7E+07 Xe-135 5.OE+07 1-134 2.2E+08 Kr-88 6.7E+07 Xe-138 1.6E+08 I-135 1.9E+08 Xe-131M 7.OE+05 Kr-83M 1.2E+07 Xe-133M 2.9E+07 5.3.3 Primary coolant 1% fuel defects activity concentration 9.11, Table 4 Isotope Activity (pCi/g)

Isotope Activity (gCi/g)

Isotope Activity ([ICi/g)

I-131 2.8E+00 Kr-85M 1.7E+00 Xe-133 2.6E+02 I-132 2.8E+00 Kr-85 8.2E+00 Xe-135M 4.9E-01 1-133 4.2E+00 Kr-87 1.OE+00 Xe-135 8.5E+00 1-134 5.7E-01 Kr-88 3.OE+00 Xe-138 6.1E-01 1-135 2.3E+00 Xe-131M 2.1E+00 Kr-83M 4.0E-01 Xe-133M 1.7E+01 5.3.4 Post-REA failed (clad 10%

9.12, page 6 damaged) fuel 5.3.5 Percent of gap activity in 9.1, Appendix H, Section 1 damaged fuel released into Containment and available for leakage from containment:

Iodine 10%

Noble Gases 10%

5.3.6 Percent of gap activity in 9.1, Appendix H, Section 1 damaged fuel released into the PC and available for P-T-S leakage:

Iodine 10%

Noble Gases 10%

5.3.7 Post-REA melted fuel 0.25%

9.12, page 6 Nuclear Common Revision 9

CALCULATION CONTINUATION SHEET SHEET 16 of 33 CALC. NO.: S-C-ZZ-MDC-1948

REFERENCE:

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

02/09/04 0

Mark Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/09/04 Design Input Parameter Value Assigned Reference 5.3.8 Percent of melted fuel 9.1, Appendix H, Section 1 released into containment and available for leakage from containment:

Iodine 25%

Noble Gases 100%

5.3.9 Percent of melted fuel 9.1, Appendix H, Section 1 released into the PC and available for P-T-S leakage:

Iodine 50%

Noble Gases 100%

5.4 Activity Transport Models (see Figures 1 and 2) 5.4.1 Containment volume 2.62 x 10' ft3 9.6.6 5.4.2 Containment leakage 0-24 hrs 0.1 w0/ofday = 0.Iv%/day 9.6.5

> 24 hrs 0.05 v%/day 9.1, Appendix H, Section 6.2 5.4.3 Containment mixing 100%

9.1, Appendix H, Section 3 5.4.4 Chemical form of iodine released to the containment Iodine Chemical Form 9.1, Appendix H, Section 4 Aerosol (CsI) 95.0%

Elemental 4.85%

Organic 0.15%

5.4.5 Steam generator dilution 119,233 Ibm per Ul Model F SG 9.7, page 7 mass 127,646.lbm per U2 Model 51 SG 9.7, page 8 5.4.6 Maximum P-T-S leakage 1 gpm (total through all four SGs) 9.6.3 & 9.6.4 rate 5.4.7 Density of P-T-S leakage 62.4 lbm/ft5 9.1, Appendix H, Section 7.2 5.4.8 Nominal reactor coolant 12,446 f' 9.6.2 system (RCS) volume 5.4.9 Steam mass released to the 5.12E+05 Ibm 9.7, page 10 environment (128,000 lbm/SG x 4 SGs) 5.4.10 Duration of steam release 110 seconds 9.7, page 10 5.4.11 Chemical form of iodine release from SGs Iodine Chemical Form 9.1, Appendix H, Section 5 Aerosol (CsI) 0%

Elemental 97%

Organic 3%

Nuclear Common Revision 9

CALCULATION CONTINUATION SHEET SHEET 17 of 33 CALC. NO.: S-C-ZZ-MDC-1948

REFERENCE:

LCR S03-05 Gopal PatelVNUCORE, ORIGINATOR, DATE REV:

02/09/04 0

Mark Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/09/04 Design Input Parameter Value Assigned Reference 5.4.12 SG liquid Iodine partition 100 9.1, Appendix H, Section 7.4 and coefficient 1 used in the analysis Appendix E, Section 5.5.4 5.5 Control Room Parameters (see Figure 3) 5.5.1 CR volume 81,420 ftl 9.10, page 33 5.5.2 CR normal flow rate 1,320 cfm (for two air intakes)

Section 6.3 5.5.3 CREACS makeup flow rate 2,200 cfm 9.6.7 5.5.4 CREACS recirc flow rate 8,000 cfm+/- 10% cfm (total) 9.6.8 with one train operating (5,000 cfm used in analysis)

Section 6.3 5.5.5 CREACS filter efficiencies Charcoal 95%

Section 6.4 HEPA 99% (95% used in the analysis) 9.6.9 and Section 6.4 5.5.6 Delay time for CR pressurization Containment Leakage Release 30 minutes Assumed Operator Action Secondary System Release 1 minute Assumed 5.5.7 CR unfiltered inleakage 150 cfm 9.15, Section 3.5.2 5.5.8 CR breathing rate 3.5E-04 me/sec 9.1, Section 4.2.6 5.5.9 CR occupancy factors Time (Hr)

Occupancy(%)

9.1, Section 4.2.6 0-24 100 24-96 60 96-720 40 5.5.10 Unit 1 CR X/Qs For Containment Leakage Release Via Unit 1 Plant Vent Time (Hr)

X/Q (sec/nd) 9.8, Section 8.2 0-2 1.78E-03 2-8 1.31E-03 8-24 5.22E-04 24-96 3.77E-04 96-720 3.17E-04 5.5.11 Unit 2 CR X/Qs For Containment Leakage Release Via Unit 1 Plant Vent Time (Hr)

X/Q (sec/nd) 9.8, Section 8.2 0-2 8.84E-04 2-8 6.60E-04 8-24 2.64E-04 24-96 1.93E-04 96-720 1.62E-04 Nuclear Common Revision 9

CALCULATION CONTINUATION SHEET SHEET 18 of 33 CALC. NO.: S-C-ZZ-MDC-1948

REFERENCE:

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

02/09/04 0

Mark Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/09/04 Design Input Parameter Value Assigned Reference 5.5.12 Unit 1 CR air intake X/Qs for Unit 1 MSSV Set 1 release Time (Hr)

X/Q (sec/n) 9.5, Section 8.1 0-2 1.57E-02 2-8 1.13E-02 8-24 4.24E-03 24-96 3.08E-03 96-720 2.26E-03 5.5.13 CR allowable dose limit 5 rem TEDE 9.3 5.5.14 CR detector specific 9.19, Table 13 efficiency Xe-133 4.1 1E7 cpm/pCi/cc Kr-85 2.51 E8 cpm/jCi/cc 5.5.15 CR intake monitor 2480 cpm 9.6.11 alarm/trip setpoint 5.6 Site Boundary Release Model Parameters 5.6.1 EAB atmospheric dispersion l 1.3E-04 l 9.9, Table 5 factor (X/Q) (sec/m3)

I 5.6.2 LPZ atmospheric dispersion factors (X/Qs)

Time (Hr)

X/Q (sec/m 3) 9.9, 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 5.6.3 EAB breathing rate (m3/sec) 3.5E-04 9.1, Section 4.1.3 5.6.4 LPZ breathing rates (m 3/sec)

Time (Hr)

BR (mplsec) 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 6.3 rem TEDE 9.1, Table 6 5.6.6 LPZ allowable dose limit 6.3 rem TEDE 9.1, Table 6 Nuclear Common Revision 9

CALCULATION CONTINUATION SHEET SHEET 19 of 33 CALC. NO.: S-C-ZZ-MDC-1948

REFERENCE:

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

02/09/04 0

l Mark Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/09/04

6.0 CALCULATIONS

Due to the limitations of the RADTRAD code, the dose consequences are determined by summing together the results from a set of three RADTRAD code analyses that separately calculate the dose contributions from the post-REA containment leakage iodine and noble gas release, secondary system iodine release originating from primary-to-secondary (P-T-S) leakage, and secondary system noble gas release originating from P-T-S leakage.

The RADTRAD RFT file SREA RFT.txt is interchangeably used for iodine release by setting the iodine release fraction to 1.0 and for the noble gas release by setting the noble gas release fraction to 1.0.

6.1 Reactor Coolant & Steam Generator Coolant Mass:

RCS Mass:

RCS water volume = 12,446 ft3 +/- 426 ft3 (Ref. 9.6.2)

RCS water nominal volume of 12,446 ft3 is used in the analysis, which yields reasonably conservative RCS activity concentrations RCS water density at cooled liquid conditions = 62.4 lb/ft3 (Ref. 9.1, Appendix H, Section 7.2)

RCS water mass = 12,446 ft3 x 62.4 lb/ft3 x 453.6 g/lbm = 3.523E+08 g SG Mass For P-T-S Iodine Activity Release:

Reference 9.7, pages 7 and 8, indicates that Salem Unit 1 has Model F steam generators each having a total liquid plus steam mass of 119,233 lbs, and Unit 2 has Model 51 steam generators each having a total mass of 127,646 lbs.

When evaluating primary-to-secondary (P-T-S) leakage, the use of a smaller secondary dilution mass results in higher secondary activity concentrations, which in turn yield higher P-T-S leakage dose consequences.

Therefore, the smaller total Unit 1 SG mass of 119,233 Ibs is used in evaluating PC iodine releases through the SGs:

Total SG volume = (4 SGs x 119,233 lb/SG) x (1/62.4) ft3/lb =

6.2 Post-REA Steaming Rates The use of specific volume of the saturated water at average steam temperature yields a higher steaming rate than the use of specific volume of saturated water at the room temperature. Therefore, the specific volume of the saturated water at average steam temperature is used in the following section to calculate the steaming rates:

6.2.1 P-T-S Leakage Iodine Release Specific volume of water ( 573 0F = 0.02253 ft3/lbm (Ref. 9.14, page 182)

Primary-to-secondary leakage = 1 gallon/minute (62.4 lb/ft3 x 0.0253 ft3/lb) x 1 ft3 = 0.188 cfm 7.482 gal Nuclear Common Revision 9 l

CALCULATION CONTINUATION SHEET SHEET 20 of 33 CALC. NO.: S-C-ZZ-MDC-1948

REFERENCE:

LCR S03-05 Gopal Patel/NUCORE,.

ORIGINATOR, DATE REV:

02/09/04 0

Mark Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/09/04 Secondary System Release Duration = I110 seconds (Ref. 9.7, page I10) 0-110 sec P-T-S leakage to SG =

p 6.2.2 SGs - Iodine Release Rate:

Mass of steam released from four SGs to the environment = 5.12E+05 Ibm (Design Input 5.4.9)

Reactor pressure vessel (RPV) average temperature = 588.40F (Ref. 9.7, page 7), which is thermal-hydraulic design value for the Model F steam generator design. The use of RPV temperature is conservative for the SG steaming rate because it yields a higher liquid density for the SG.

SG Vessel average temperature = 588.40F (Ref. 9.7, page 7) is conservatively used to maximize the steaming rate because the densities at the SG pressures 823.4 psia/768 psia (Ref. 9.7, page 8)) would be 0.021007 ft3/lb/0.02081 ft3/lb, which are lower than 0.02313 ft3/lb used in the following section.

Specific volume of liquid = 0.02313 ft3/lb (Ref. 9.14, page 183)

Iodine partition coefficient = 1.0 (no credit taken for P-T-S leakage iodine retention in the SGs liquid) 0-110 sec SG steaming rate = 5.12E+05 x 0.02313 ft3/lb x (1/1.0) =

(110 sec)/(60 min/sec) 6.2.3 SGs - Noble Gas Release Rate:

No noble gas hold-up or decay is modeled in the four SGs. All RCS mass released to the SGs is treated as if it were released directly to the environment at the primary-to-secondary leakage rate calculated in Section 6.2.1.

0-110 seconds noble gas release rate to the environment = IM,k0 0

S 1

6.3 CREACS Air Flow Rates:

Normal Flow Rate Notes "S" to the Reference 9.16 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 I 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 I (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 c~ft Per Notes "S", use 600 cfm for Zone 1 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 CREACS Recirculation Flow Rate With CR Monitors Preferentially Selecting Less Contaminated Air Intake CREACS ventilation flow rate = 8,000 cfm +/- 10% cfm (Ref. 9.6.8)

Minimum CREACS flow rate = 8,000 cfm - (0.10 x 8,000 cfm) = 8,000 cfm - 800 cfm = 7,200 cfm l Nuclear Common Revision 9 ]

I CALCULATION CONTINUATION SHEET ISHEET 21 of 33 CALC.NO.: S-C-ZZ-MDC-1948

REFERENCE:

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

02/09/04 0

Mark Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/09/04 Net CREACS recirculation flow rate = Minimum CREACS flow rate - CREACS makeup flow rate 7,200 cfm - 2,200 cfm (Ref. 9.6.7) = 5,000 cfm 6.4 CREACS Charcoal/HEPA Filter Efficiencies:

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. 9.6.10)

GL 99-02 (Ref. 9.17) requires a safety factor of at least 2 should be used to determine the filter efficiencies to be credited in the design basis accident.

Testing methyl iodide penetration (%/) = (100% - rj)/safety factor = (100% - tl)/2 Where Tj = charcoal filter efficiency to be credited in the analysis CREACS Charcoal Filter 2.5% = (100% - il)/2 5%= (100%-ni) q = 100% - 5%= 95%

HEPA Filter HEPA filter efficiency = 99% (Ref. 9.6.9). HEPA filter efficiency of 95% is used in the analysis Safety Grade Filter Efficiency Credited (%)

Filter Aerosol I

Elemental I

Organic CREACS 95 1

95.

95 6.5 CR Intake Monitor Setpoint Evaluation:

CR Detector Specific Efficiencies:

Xe-133 = 4.11E7 cpm/ljCi/cc (Ref. 9.19, Table 13)

Kr-85 = 2.5 1E8 cpm./lCi/cc (Ref. 9.19, Table 13)

There are two post-REA release paths to the environment - containment leakage path and P-T-S leakage path via steam generator.

Containment Leakage:

Concentration = (Containment activity based on failed fuel)/(Cont. volume) x (Cont. leak rate) x (X/Q)

Containment volume = 2.62E+06 ft3 (Ref. 9.6.6)

Containment leak rate = 0.1 V%/day for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (Design Input 5.4.2).

Release rate through cont. leakage at start of REA = (2.62E+06 ft3 x 0.001 day-')/(1440 min/day) = 1.82 cfm I Nuclear Common Revision 9 1 Nuclear Common Revision 9

CALCULATION CONTINUATION SHEET SHEET 22 of 33 CALC. NO.: S-C-ZZ-MDC-1948

REFERENCE:

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

02/09/04 0

Mark Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/09/04 The initial Xe-133 and Kr-85 concentrations at the CR intake for the design basis REA are calculated as follows:

Containment Activity Xe-133 = 2.614E+06 Ci and Kr-85 = 1.676E+04 Ci (Table 3)

Containment Volume = 2.62E+06 ft3 (Ref. 9.6.6) and maximum 0-2 hour plant vent X/Q = 1.78E-03 sec/m3 (Ref. 9.5, Section 8.2)

Xe-133 Concentration at the CR intake:

= (2.614E+06 Ci/2.62E+06 ft3) x (1.82 ft3/rnin) x (1.78E-03 sec/m 3) x (1/60 min/sec)

= 5.39E-05 Ci/m3 = 5.39E-05 pCi/cc Kr-85 Concentration at the CR intake:

= (1.676E+04 Ci/2.62E+06 ft3) x (1.82 ft3/min) x (1.78E-03 sec/m3) x (1/60 min/sec)

= 3.45E-07 Ci/m3 = 3.45E-07 pCi/cc Count Rate Xe-133 = 5.39E-05 pCi/cc x 4.1 1E7 cpm/pCi/cc = 2,215 cpm Count Rate Kr-85 = 3.45E-07 pCi/cc x 2.51E8 cpm/pCi/cc = 86 cpm Combined Count Rate of CR Intake Detector

= 2,215 cpm + 86 cpmr = 2,301 cpm < CR Monitor Setpoint of 2,480 cpm Therefore, the containment release would not be detected.

P-T-S Leakage Path Concentration = (RCS activity based on failed fuel)/( RCS volume) x (P-T-S leak rate) x (X/Q)

RCS Activity Xe-133 = 2.614E+06 Ci and Kr-85 = 1.676E+04 Ci (Table 4)

RCS Volume = 12,446 ft3 (Ref. 9.6.2) and 0-2 hour X/Q = 1.57E-02 sec/m 3 (Ref. 9.5, Section 8.1)

RCS Leak Rate of Noble Gas isotopes directly to the environment

= 0.188 ft3/rnin/4 = 0.047 ft3/min) (Section 6.2.3) (Assuming one intact SG releases through the MSSV Set 1 and other intact SG are considered in the setpoint evaluation)

Xe-133 Concentration at the CR intake

= (2.614E+06 Ci/12,446 ft3) x (0.047 ft3/min) x (1.57E-02 sec/m3) x (1/60 min/sec)

= 2.58E-03 Ci/m3 = 2.58E-03 pCi/cc Kr-85 Concentration at the CR intake

= (1.676E+04 Ci/12,446 ft3) x (0.047 ft3/min) x (1.57E-02 sec/m3) x (1/60 min/sec)

= 1.66E-05 Ci/m3 = 1.66E-05 plCi/cc Count Rate Xe-133 = 2.58E-03 gCi/cc x 4.1 1E7 cpm/pCi/cc = 106,038 cpm Count Rate Kr-85 = 1.66E-05 pCi/cc x 2.51E8 cpm/lpCi/cc = 4,167 cpm Combined Count Rate of CR Intake Detector = 106,038 cpm + 4,167 cpm I Nuclear Common Revision 9 1 Nuclear Common Revision 9

I CALCULATION CONTINUATION SHEET ISHEET 23 of 33 CALC. NO.: S-C-ZZ-MDC-1948

REFERENCE:

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

02/09/04 0

l Mark Drucker/NUCORE, REVIEWERNVERIFIER, DATE 02/09/04

= 110,205 cpm >> CR Monitor Setpoint of 2,480 cpm The Post-REA concentration at the CR intake due to P-T-S leakage greatly exceeds the CR monitor setpoint value of 2,480 cpm, therefore the CR monitor will respond instantaneously.

7.0 RESULTS

SUMMARY

The results of the Rod Ejection accident are summarized in the following table with the CR monitors preferentially selecting the less contaminated air intake:

Rod Ejection Accident TEDE Dose (rem)

Receptor Location Control Room EAB LPZ Containment Leakage 1.03E+00 2.15E-01 1.26E-01 SREACLO1 (occurs @

0.0 hr)

P-T-S Iodine Release 3.26E-01 3.50E-02 5.OOE-03 SREAID00 (occurs @ 0.0 hrs)

Noble Gas Release 7.41E-03 3.14E-03 4.50E-04 SREANG00 (occurs ( 0.0 hr)

Total 1.36E+00 2.53E-01 1.31E-01 Allowable TEDE Limit 5.OOE+00 6.30E+00 6.30E+00

8.0 CONCLUSION

S:

The Rod Ejection Accident results presented in Section 7.0 indicate that the EAB, LPZ, and CR doses due to a REA are within their allowable limits.

Nuclear Common Revision 

I Nuclear Common Revision 9 1

CALCULATION CONTINUATION SHEET SHEET 24 of 33 CALC. NO.: S-C-ZZ-MDC-1948

REFERENCE:

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

02/09/04 0

Mark Drucker/NUCORE, REVIEWERIVERIFIER, DATE 02/09/04

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.

10 CFR 50.67, "Accident Source Term."

4.

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

5.

Calculation No. S-C-ZZ-MDC-1 959, Rev. 0, CR x/Qs Using ARCON96 Code - Non-LOCA Releases

6.

Salem 1 & 2 Technical Specifications:

1.

Specification 1.25, Salem Unit 1/Unit 2 Rated Thermal Power

2.

Specification 5.4.2, Salem Unit 1/Unit 2 Reactor Coolant System Volume

3.

Specification 3.4.6.2, Salem Unit 1 Limiting Condition for Operation (LCO) for Reactor Coolant System Operational Leakage

4.

Specification 3.4.7.2, Salem Unit 2 LCO for Reactor Coolant System Operational Leakage

5.

Specification 6.8.4.f, Primary Containment Leakage Rate Testing Program -

6.

Specification 5.2.1, Containment Configuration

7.

Specification Surveillance Requirement 4.7.6.1.d.3, Salem Unit 1/Unit 2 CREACS Design Makeup Flow Rate

8.

Specification Surveillance Requirement 4.7.6.1.d.1, Salem Unit 1/Unit 2 CREACS Ventilation Flow Rate

9.

Specification Surveillance Requirement 4.7.6.1.e, Salem Unit 1/Unit 2 HEPA Filter DOP

10.

Specification Surveillance Requirement 4.7.6.1.b.3, Salem Unit 1/Unit 2 CREACS Methyl Iodide Penetration

11.

Specification 3.3.3.1 and Table 3.3-6, Salem Unit 1/Unit 2 LCO for Radiation Monitoring Instrumentation

12.

Specification 3.7.6.1, Salem Unit 1 LCO for Control Room Emergency Air Conditioning System

13.

Specification 3.7.6, Salem Unit 2 LCO for Control Room Emergency Air Conditioning System

7.

Nuclear Fuel Section Calculation File No. DS 1.6-0453, Determination of Steam Release Flows for Input to the Radiological Dose Analysis

8.

SNGS Calculation No. S-C-ZZ-MDC-1912, Rev. 0, CR X/Qs Using ARCON96 Code - Equipment Hatch, Plant Vent, & FHB Rollup Door Releases I Nuclear Common Revision 9 1 I Nuclear Common Revision 9 I

I CALCULATION CONTINUATION SHEET ISHEET 25 of 33 CALC. NO.: S-C-ZZ-MDC-1948

REFERENCE:

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

02/09/04 0

l Mark Drucker/NUCORE, REVIEWER(VERIFIER, DATE 02/09/04

9.

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

10.

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

11.

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

12.

Westinghouse Calculation No. CN-CRA-93-105, Rev. 0, Salem Control Rod Ejection Accident Offsite Dose

13.

Safety Evaluation By The Office of Nuclear Reactor Regulation Related to Amendment Nos. 257 & 139 To Facility Operating License Nos. DPR-66 and NPF-73 for Beaver Valley Power Station Unit Nos. 1 and 2 Docket Nos. 50-334 and 50-412 (TAC Nos. MB5303 and MB5304)

14.

ASME Steam Tables, Sixth Edition

15.

SNGS Calculation No. S-C-ZZ-MDC-1 987, Rev. 1, Input Parameters for Salem AST Dose CaIcs

16.

SNGS Mechanical P&IDs:

a.

205248, Sheet 2, Rev. 44, Unit 1 Aux Bldg Control Area Air Conditioning & Ventilation

b.

205348, Sheet 2, Rev. 34, Unit 2 Aux Bldg Control Area Air Conditioning & Ventilation

17.

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

18.

Not Used.

19.

Vendor Technical Document No. 311649, Rev. 1, Accuracy Analysis Of The Sorrento Electronics WRGM, Liquid Effluent, And In-Line Duct Monitors.

20.

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

21.

Federal Guidance Report 12, EPA-402-R-93-081, Environmental Protection Agency INuclear Common Revision 9 7 Nuclear Common Revision 9 I

I CALCULATION CONTINUATION SHEET ISHEET 26 of 33 CALC. NO.: S-C-ZZ-MDC-1948

REFERENCE:

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

02/09/04 0

Mark Drucker/NUCORE, REVIEWERIVERIFIER, DATE 02/09/04 10.0 TABLES:

Table 1 Iodine & Noble Gas Design Basis Core Inventory Core Inventory Core Design Basis Based on Rated Core Isotope 3600 MWt Thermal Inventory Power (Ci)

(MWO (Ci)

A B

AxBxl.05/3600 KR-83M 1.200E+07 3459 1.21 IE+07 KR-85 1.100E+06 3459 1.1I OE+06 KR-85M 2.600E+07 3459 2.623E+07 KR-87 4.700E+07 3459 4.742E+07 KR-88 6.700E+07 3459 6.759E+07 1-131 9.900E+07 3459 9.988E+07 1-132 1.400E+08 3459 1.412E+08 1-133 2.OOOE+08 3459 2.018E+08 1-134 2.200E+08 3459 2.220E+08 I-135 1.900E+08 3459 1.917E+08 XE-131M 7.OOOE+05 3459 7.062E+05 XE-133 2.OOOE+08 3459 2.018E+08 XE-133M 2.900E+07 3459 2.926E+07 XE-135 5.000E+07 3459 5.044E+07 XE-135M 4.OOOE+07 3459 4.036E+07 XE-138 1.600E+08 3459 1.614E+08 A From Reference 9.11, Table 2 B From Reference 9.6.1 I Nuclear Common Revision 9 1 Nuclear Common Revision 9

CALCULATION CONTINUATION SHEET SHEET 27 of 33 CALC. NO.: S-C-ZZ-MDC-1948

REFERENCE:

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

02/09/04 0

Mark Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/09/04 Table 2 Total Iodine & Noble Gas Activity in RCS 1% Fuel PC PC Defects Mass In Total Isotope PC Activity RCS Activity Concentration (WCitg)

(g)

(Ci)

A B

C=AxB/lE+6 KR-83M 4.000E-01 3.523E+08 1.409E+02 KR-85 8.200E+00 3.523E+08 2.889E+03 KR-85M 1.700E+00 3.523E+08 5.989E+02 KR-87 1.OOOE+00 3.523E+08 3.523E+02 KR-88 3.OOOE+00 3.523E+08 1.057E+03 I-131 2.80E+00 3.523E+08 9.864E+02 1-132 2.80E+00 3.523E+08 9.864E+02 I-133 4.20E+00 3.523E+08 1.480E+03 1-134 5.70E-01 3.523E+08 2.008E+02 1-135 2.30E+00 3.523E+08 8.103E+02 XE-13 IM 2.100E+00 3.523E+08 7.398E+02 XE-133 2.600E+02 3.523E+08 9.160E+04 XE-133M 1.700E+01 3.523E+08 5.989E+03 XE-135 8.500E+00 3.523E+08 2.995E+03 XE-135M 4.900E-01 3.523E+08 1.726E+02 XE-138 6.100E-01 3.523E+08 2.149E+02 A From Reference 9.11, Table 4 B From Section 6.1 Nuclear Common Revision 9 I Nuclear Common Revision 9 l

CALCULATION CONTINUATION SHEET SHEET 28 of 33 CALC. NO.: S-C-ZZ-MDC-1948

REFERENCE:

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

02/09/04 0

Mark Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/09/04 Table 3 Post-REA Iodine & Noble Gas Activity Released To Containment Design Basis Cont.

Cont.

Cont.

Total RADTRAD Core Activity Activity Activity Activity Nuclide Isotope Inventory From From From Released Inventory 10%

0.25%

Primary To File Failed Fuel Melted Fuel Coolant Containment

,(Ci)

(Ci)

(Ci)

(Ci)

(Ci)

(Ci)

A B

C D

E=B+C+D KR-83M 1.21 1E+07 1.211E+05 3.027E+04 1.409E+02 1.515E+05

.1515E+06 KR-85 1.11 OE+06 1.110E+04 2.774E+03 2.889E+03 1.676E+04

.1676E+05 KR-85M 2.623E+07 2.623E+05 6.558E+04 5.989E+02 3.285E+05

.3285E+06 KR-87 4.742E+07 4.742E+05 1.185E+05 3.523E+02 5.931E+05

.5931E+06 KR-88 6.759E+07 6.759E+05 1.690E+05 1.057E+03 8.460E+05

.8460E+06 1-131 9.988E+07 9.988E+05 6.242E+04 9.864E+02 1.062E+06

.1062E+07 1-132 1.412E+08 1.412E+06 8.828E+04 9.864E+02 1.502E+06

.1502E+07 1-133 2.018E+08 2.018E+06 1.261E+05 1.480E+03 2.145E+06

.2145E+07 I-134 2.220E+08 2.220E+06 1.387E+05 2.008E+02 2.358E+06

.2358E+07 1-135 1.917E+08 1.917E+06 1.198E+05 8.103E+02 2.037E+06

.2037E+07 XE-13 IM 7.062E+05 7.062E+03 1.766E+03 7.398E+02 9.567E+03

.9567E+04 XE-133 2.018E+08 2.018E+06 5.044E+05 9.160E+04 2.614E+06

.2614E+07 XE-133M.

2.926E+07 2.926E+05 7.314E+04 5.989E+03 3.717E+05

.3717E+06 XE-135 5.044E+07 5.044E+05 1.261E+05 2.995E+03 6.335E+05

.6335E+06 XE-135M 4.036E+07 4.036E+05 1.009E+05 1.726E+02 5.046E+05

.5046E+06 XE-138 1.614E+08 1.614E+06 4.036E+05 2.149E+02 2.018E+06

.2018E+07 A From Table 1 B = (A x 0.10 x 0.10) = 0.01A for iodine [10% iodine release from 10% clad damage]

= (A x 0.10 x 0.10) = 0.01A for noble gas [10% noble gas release from 10% clad damage]

C = (A x 0.25 x 0.0025) = 0.000625A for iodine [25% iodine release from 0.25% melted fuel]

= (A x 1.00 x 0.0025) = 0.0025A for noble gas [100% noble gas release from 0.25% melted fuel]

D From Table 2 I Nuclear Common Revision 9 1 Nuclear Common Revision 9

CALCULATION CONTINUATION SHEET SHEET 29 of 33 CALC. NO.: S-C-ZZ-MDC-1948

REFERENCE:

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

02/09/04 0

Mark Drucker/NUCORE, REVIEWERIVERIFIER, DATE 02/09/04 Table 4 Post-REA Iodine & Noble Gas Activity Released To RCS For P-T-S Leakage Design Basis RCS RCS Initial Total RADTRAD Core Activity Activity Activity Activity Nuclide Isotope Inventory From From In Released Inventory 10%

0.25%

Primary To File Failed Fuel Melted Fuel Coolant RCS (Ci)

(Ci)

(Ci)

(Ci)

(Ci)

(Ci)

A B

C D

E=B+C+D KR-83M 1.21 IE+07 1.21 1E+05 3.027E+04 1.409E+02 1.515E+05

.1515E+06 KR-85 1.111OE+06 1.1 IOE+04 2.774E+03 2.889E+03 1.676E+04

.1676E+05 KR-85M 2.623E+07 2.623E+05 6.558E+04 5.989E+02 3.285E+05

.3285E+06 KR-87 4.742E+07 4.742E+05 1.185E+05 3.523E+02 5.931E+05

.5931E+06 KR-88 6.759E+07 6.759E+05 1.690E+05 1.057E+03 8.460E+05

.8460E+06 1-131 9.988E+07 9.988E+05 1.248E+05 9.864E+02 1.125E+06

.1125E+07 1-132 1.412E+08 1.412E+06 1.766E+05 9.864E+02 1.590E+06

.1590E+07 1-133 2.018E+08 2.018E+06 2.522E+05 1.480E+03 2.271E+06

.2271E+07 1-134 2.220E+08 2.220E+06 2.774E+05 2.008E+02 2.497E+06

.2497E+07 1-135 1.917E+08 1.917E+06 2.396E+05 8.103E+02 2.157E+06

.2157E+07 XE-131M 7.062E+05 7.062E+03 1.766E+03 7.398E+02 9.567E+03

.9567E+04 XE-133 2.018E+08 2.018E+06 5.044E+05 9.160E+04 2.614E+06

.2614E+07 XE-133M 2.926E+07 2.926E+05 7.314E+04 5.989E+03 3.717E+05

.3717E+06 XE-135 5.044E+07 5.044E+05 1.261E+05 2.995E+03 6.335E+05

.6335E+06 XE-135M 4.036E+07 4.036E+05 1.009E+05 1.726E+02 5.046E+05

.5046E+06 XE-138 1.614E+08 1.614E+06 4.036E+05 2.149E+02 2.018E+06

.2018E+07 A From Table I B = (A x 0.10 x 0.10) = 0.01A for iodine [10% iodine release from 10% clad damage]

= (A x 0.10 x 0.10) = 0.0IA for noble gas [10% noble gas release from 10% clad damage]

C = (A x 0.50 x 0.0025) = 0.00125A for iodine [50% iodine release from 0.25% melted fuel]

= (A x 1.00 x 0.0025) = 0.0025A for noble gas [100% noble gas release from 0.25% melted fuel]

D From Table 2 I Nuclear Common Revision 9

l CALCULATION CONTINUATION SHEET ISHEET 30 of 33 CALC. NO.: S-C-ZZ-MDC-1948

REFERENCE:

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

02/09/04 0

l l

Mark Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/09/04 11.0 FIGURES:

Figure 1: RADTRAD Nodalization For Post-REA Containment Leakage Release I Nuclear Common Revision 9 l Nuclear Common Revision 9 I

I CALCULATION CONTINUATION SHEET ISHEET 31 of 33 CALC. NO.: S-C-ZZ-MDC-1948

REFERENCE:

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

02/09/04 0

Mark Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/09/04 Figure 2: RADTRAD Nodalization For Post-REA P-T-S Iodine & Noble Gas Releases P-T-S Leakage Iodine Release P-T-S Leakage Noble Gas Release I Nuclear Common Revision 9 1 Nuclear Common Revision 9 I

I CALCULATION CONTINUATION SHEET I SHEET 32 of 33 CALC. NO.: S-C-ZZ-MDC-1948

REFERENCE:

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

02/09/04 0

l Mark Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/09/04 Figure 3 - RADTRAD Nodalization for Post-REA CR Response t = 30 minutes for containment leakage release t = 1 minute for secondary system release Nuclear Common Revision 9 I Nuclear Common Revision 9 l

I CALCULATION CONTINUATION SHEET ISHEET 33 of 33 CALC. NO.: S-C-ZZ-MDC-1948

REFERENCE:

LCRS03-05 Gopal Patel/NUCORE, ORIGINATOR, DATE REV:

02/09104 0

l Mark Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/09/04 12.0 AFFECTED DOCUMENTS:

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

Document to be'voided:

Vendor Technical Document No. 322261, Rev 3, Radiological Dose Consequence at the EAB/LPZ and in the Control Room with Modified Control Room Ventilation Design - Control Rod Ejection Accident.

13.0 ATTACHMENTS:

One Diskette with the following electronic files (1 page):

Calculation No: S-C-ZZ-MDC-1948, Rev OIR2 Comment Resolution Form 2 - Mark Drucker Certification for Design Verification Form-1 RCPD Form-l INuclear Common Revision 9 l Nuclear Common Revision 9

Attachment A S-C-ZZ-MDC-1948, Rev. 0 1 Diskette With Various Electronic Files