ML042740520

Eab, LPZ, & CR Doses - RCP Locked Rotor Accident (LRA) - AST

ML042740520
Person / Time
Site:	Salem
Issue date:	02/05/2004
From:	Gita Patel NUCORE
To:	Office of Nuclear Reactor Regulation
References
LR-N04-0413 S-C-ZZ-MDC-1951, Rev OIR1
Download: ML042740520 (35)

Text

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

CALC NO.: S-C-ZZ-MDC-1951 CALCULATION COVER SHEET Page 1 of 34 REVISION: CIR1 CALC. TITLE:

EAB, LPZ, & CR Doses - RCP Locked Rotor Accident (LRA) - AST

1. SHTS (CALC):

4 ATT I # SHTS:

1/1

1. IDV/50.59 SHTS: I 213

1. TOTAL SHTS:

40 CHECK ONE:

R FINAL 0 INTERIM (Proposed Plant Change)

El FINAL (Future Confirmation Req'd) a VOID SALEM OR HOPE CREEK:

E Q - LIST 0 IMPORTANT TO SAFETY ED NON-SAFETY RELATED HOPE CREEK ONLY:

FlQ FlQs ElQsh OIF EOR 0D STATION PROCEDURES IMPACTED, IF SO CONTACT RELIABILITY ENGINEER El CDs INCORPORATED (IF ANY):

DESCRIPTION OF CALCULATION REVISTON (IF APPL.Q:

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 Reactor Coolant Pump (RCP) Locked Rotor Accident 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 results of this analysis, as presented in Section 7, indicate that the EAB, LPZ, and CR doses due to a RCP locked rotor accident are within their allowable TEDE dose limits.

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l Mark Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/05/2004 REVISION HISTORY Revision Revision Description OIRO Initial Issue.

OIRI Design validation comments incorporated.

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Mark Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/05/2004 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 18 7.0 Results Summary 22 8.0 Conclusions 22 9.0 References 23 10.0 Tables 25 11.0 Figures 31 12.0 Affected Documents 34 13.0 Attachments 34 Nuclear Common Revision 91

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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 Reactor Coolant Pump (RCP) Locked Rotor Accident (LRA) 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 LRA 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 LRA is analyzed using plant specific design and licensing bases inputs which are compatible to the TEDE dose criteria. The LRA analysis is performed using the guidance in Regulatory Guide 1.183. and its Appendix G (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 LRA. 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 Salem 1 & 2 plants (Ref. 9.19). 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 ducts equipped with safety related fans and radiation monitors (Ref. 9.19), which make the CREACS air supply system single failure proof. The redundant air intake monitors preferentially select the less contaminated air intake during an accident condition when the radiation level at the normal outside air intake exceeds the setpoint value (see Section 6.5 for 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, REVIEWER/VERIFIER, DATE 02/05/2004 The principal radiation source term for the LRA is the fuel gap activity in the failed fuel caused by the transient of the accident. It is assumed that 5% of the fuel cladding in the core has failed, with the failed fuel rod gap activity released immediately to the primary coolant (PC) (Ref. 9.13, page 4). The fuel gap isotopic activity release fractions are in accordance with RG 1.183 (Ref. 9.1, Table 3). Previous LRA analysis had assumed a preaccident iodine spike activity release in addition to the failed fuel activity release (Ref. 9.13, page 4). In this analysis the preaccident iodine spike activity release is not modeled. The exclusion of the preaccident iodine spike is consistent with the guidance presented in RG 1.183 (Ref. 9.1, Appendix E, Footnote 2), which states that the activity assumed in the analysis should be based on the activity associated with the projected fuel damage or the maximum technical specification values, whichever maximizes the radiological consequences.

Per the comparison of 5% failed fuel and preaccident iodine spike activities presented on page 6 of Reference 9.13, an activity release based on the projected 5% fuel damage would yield higher dose consequences than would a release of activity at the maximum technical specification limits. Therefore, the preaccident iodine spike activity release is not modeled with the failed fuel activity release.

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 design basis core iodine and noble gas activity releases from the 5% failed fuel are calculated in Table 2 using the iodine and noble gas gap release fractions addressed in RG 1.183, Table 3. The activity released from the failed fuel is assumed to be instantaneously and homogeneously mixed through the primary coolant. The initial iodine and noble gas activity in the PC is calculated in Table 3 using the 1% fuel defects RCS activity concentrations and the Reactor Coolant System (RCS) coolant mass. This initial PC activity is added to the 5%

failed fuel activity in Table 4, which is used to develop RADTRAD Nuclide Inventory File (NIF)

SLRAldef.txt. The RCS activity is assumed to be released into the four SGs via a primary-to-secondary (P-T-S) leak rate of 1 gpm (Refs. 9.6.3 & 9.6.4). Noble gases present in the P-T-S leakage are released directly to the environment without retention in the SGs. It is assumed that the SG tubes are 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.

Thirty-two hours after the accident, the RHR system is assumed to be in operation (Ref. 9.18) and releases from the SGs have been terminated (Ref. 9.1, Appendix G, Section 5.3). The steaming rates of SGs are calculated based on the loss of load mass releases, which are applicable to the locked rotor accident (Ref. 9.21).

The initial secondary coolant iodine activity used in this calculation is based on the technical specification limit of 0.1 jiCi/g Dose Equivalent of I-131 (Ref. 9.6.5). The secondary liquid iodine activity is calculated in Tables 5 through 8, and input into RADTRAD NIF file SLRASLIodine def.txt to calculate dose consequences due to the secondary liquid iodine activity release. Secondary liquid iodine is assumed to be directly released to the environment in proportion to the steam release rate, with credit taken for iodine partitioning in the SG liquid.

For the secondary liquid iodine release from the SGs, an iodine partition coefficient of 10 is conservatively assumed in lieu of the value of 100 recommended in Regulatory Guide 1.183 (Ref. 9.1, Appendix G, Section 5.6 and Appendix E, Section 5.5.4). Thirty-two hours after the accident, the RHR system is assumed to be in operation (Ref. 9.18), and no steam is released after this time.

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Mark Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/05t2004 The doses from the LRA releases - iodine releases from the P-T-S leakage, iodine releases from the secondary liquid, and noble gas releases from the P-T-S leakage - are summarized in Section 7.0.

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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 LRA. The following sections address the applicability of these modeling assumptions to an SNGS Units 1 and 2 Locked Rotor accident analysis. These assumptions are incorporated as design inputs in Section 5.3.

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 There is 5% fuel damage postulated for the LRA (Ref. 9.13, page 4). Per RG 1.183 (Ref. 9.1, Appendix G, Section 2), since fuel damage is postulated for the limiting event, a radiological analysis is required.

4.1.1 Activity Release from the Fuel into the Primary Coolant Consistent with RG 1.183 (Ref. 9.1, Appendix G, Section 3) the activity released from the fuel is assumed to be released instantaneously and homogeneously through the primary coolant.

4.1.2 Iodine Chemical Form Consistent with RG 1.183 (Ref. 9.1, Appendix G, Section 4) the chemical form of iodine releases from the steam generators to the environment are assumed to be 97% elemental and 3% organic. These fractions apply to iodine releases from the primary-to-secondary (P-T-S) leakage and the secondary liquid.

4.2 Transport 4.2.1 Primary-to-Secondary Leak Rates Consistent with RG 1.183 (Ref. 9.1, Appendix G, Section 5.1) the primary-to-secondary leak rate in the four steam generators is assumed to be the leak rate limn b

for operation specified in the technical specifications (Refs. 9.6.3 and 9.6.4).

4.2.2 Primary-to-Secondary Leak Primary Coolant Density Consistent with RG 1.183 (Ref. 9.1, Appendix G, Section 5.2) 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 Ibm/fl3). It is assumed that the SG tubes are not submerged in the SGs liquid, and consequently iodine introduced into the four SGs via the P-T-S leakage is assumed to be directly released to environment with no credit taken for iodine partitioning in the SG liquid. For the secondary liquid iodine release from the Nuclear Common Revision 9 I INuclear Common Revision 91l

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SGs, an iodine partition coefficient of 10 is conservatively assumed in lieu of the value of 100 recommended in Regulatory Guide 1.183 (Ref. 9.1, Appendix G, Section 5.6 and Appendix E, Section 5.5.4).

4.2.3 Primary-to-Secondary Leak Duration Consistent with RG 1.183 (Ref. 9.1, Appendix G, Section 5.3) the primary-to-secondary leakage is assumed to continue until the primary system pressure is less than the secondary system pressure, or until the temperature of the leakage is less than 212'F. In this analysis primary coolant is conservatively assumed to leak into the SGs for 32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br /> until the RHR system is initialized (Ref. 9.18).

4.2.4 Release of Fission Products Consistent with RG 1.183 (Ref. 9.1, Appendix G, Section 5.4) the release of fission products from the secondary system is evaluated with the assumption of a coincident loss of offsite power (LOOP). The offsite power is assumed to be lost so that the main steam condensers are not available for removal of the decay heat.

4.2.5 Noble Gas Releases to the Environment Consistent with RG 1.183 (Ref. 9.1, Appendix G, Section 5.5) all noble gas radionuclides released from the primary system are assumed to be released to environment without reduction or mitigation.

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 LRA analysis.

4.3.1 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" 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. 9.7).

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l Mark Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/05/2004 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 assumed to be 2.3 x 104 cubic meters per second. These offsite breathing rates are listed in Sections 5.6.3 &

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. 9.8).

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 CFR 50.67 (Ref. 9.3).

For the LRA the postulated EAB doses should not exceed the criteria established in RG 1.183 Table 6:

EAB Dose Acceptance Criterion:

2.5 Rem TEDE Per RG 1.183 Table 6, the LRA event release duration is until cold shutdown is established.

The RADTRAD Code (Ref. 9.2) 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 LRA the postulated LPZ doses should not exceed the criteria established in RG 1.183 Table 6:

LPZ Dose Acceptance Criterion:

2.5 Rem TEDE Per RG 1.183 Table 6, the LRA event release duration is until cold shutdown is established.

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Mark Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/05/2004 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.

4.4 Control Room Offsite 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 LRA analysis. These assumptions are incorporated as design inputs in Sections 5.5.1 through 5.5.14.

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). This form of radiation shine is not applicable to a LRA in which the activity is not released into the containment air space, 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 and the CR filter shine dose due to a LRA are insignificant compared to those due to a LOCA (see the core release fractions for LOCA and non-LOCA design basis accidents in Tables I and 3 of Reference 9.1). Therefore, these direct dose contributions are considered to be insignificant and are not evaluated for a LRA.

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.

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Mark Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/05/2004 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).

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.

4.4.7 Control Room Dose Conversion Factors Consistent with RG 1.183 (Ref. 9.1, Section 4.2)75TifEcontrol room doses are calculated using the offsite dose analysis dose conversion factors identified in RG.1;83 Regulatory Position 4.1 (see above Sections 4.3.2 &

4.3.4). The deep dose equivalent (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, DIErinite, 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 V0. 338) / 1173 I Nuclear Common Revision 91 I Nuclear Common Revision 9 I

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Mark Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/05/2004 4.4.8 Control Room Operator Dose Acceptance Criteria Consistent with RG 1.183 (Ref. 9.1, Section 4.4), for the LRA the postulated CR doses should not exceed the 5 Rem TEDE criterion established in 10 CFR 50.67 (Ref 9.4):

CR Dose Acceptance Criterion:

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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-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 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 the assumptions are consistent with those identified in Appendix G of RG 1.183 (Ref. 9.1). The design inputs and assumptions in the following sections represent the as-built design of the plant.

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Mark Drucker/NUCORE, REVIEWERtVERIFIER, DATE 02/05/2004 Design Input Parameter Value Assigned Reference Locked Rotor 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. IE+06 Xe-135M 4.0E+07 I-133 2.OE+08 Kr-87 4.7E+07 Xe-135 5.OE+07 I-134 2.2E+08 Kr-88 6.7E+07 Xe-138 1.6E+08 1-135 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 (IiCilg)

Isotope Activity (gCi/g)

Isotope Activity (.Ci/g) 1-131 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 I-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 PC Tech Spec activity 9.6.10 & 9.6.11 Iodine 1.0 piCi/gm DE Iodine-131 Non-Iodine (i.e., Noble Gas) 100 / E-BAR 5.3.5 Post-LRA failed fuel 5%

9.13, page 4 5.3.6 Fission product inventory in 9.1, Table 3 Fuel Rod Gap:

1-131 0.08 Other Halogens (i.e., iodine) 0.05 Kr-85 0.10 Other Noble Gases 0.05 5.3.7 Secondary coolant (iodine) 0.1 I Ci/gm DE Iodine-131 9.6.5 specific activity in the Technical Specifications Nuclear Common Revision 9

CALCULATION CONTINUATION SHEET SHEET 16 of 34 CALC. NO.: S-C-ZZ-MDC-1951

REFERENCE:

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

02/05/2004 0

Mark Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/05/2004 Design Input Parameter Value Assigned Reference 5.4 Activity Transport Models (Figures 1 & 2) 5.4.1 Loss of load steam release 9.21, page 8 0-2 hours 6.55E+05 lbs 2-8 hours 5.40E+05 lbs 8-32 hours 2.40E+06 lbs 5.4.2 SG liquid Iodine partition 100 9.1, Appendix G, Section 5.6 and coefficient (10 used in the analysis)

Appendix E, Section 5.5.4 5.4.3 Primary-to-secondary leak 32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br /> 9.18 rate duration 5.4.4 Nominal reactor coolant 12,446 ft3 9.6.2 system (RCS) volume 5.4.5 Total P-T-S leakage through 1 gpm 9.6.3 & 9.6.4 all steam generators (SGs) 5.4.6 Steam generator dilution 119,233 lbm per Ul Model F SG 9.16, page 7 mass 127,646 Ibm per U2 Model 51 SG 9.16, page 8 5.5 Control Room Model Parameters (Figure 3) 5.5.1 CR volume 81,420 ft 9.10, page 33 5.5.2 CR normal flow rate 1,320 cfhi (for two air intakes)

Section 6.3 5.5.3 CREACS makeup flow rate 2,200 cfm 9.6.6 5.5.4 CREACS recirc flow rate 8,000 cfm +/- 10% cfm 9.6.7 with one train operating (5,000 cfln used in analysis)

Section 6.3 5.5.5 CREACS filter efficiencies Charcoal 95%

Section 6.4 HEPA 99% (95% used in the analysis) 9.6.8 and Section 6.4 5.5.6 Delay time for CR 1 minute Assumed pressurization 5.5.7 CR unfiltered inleakage 150 cfm 9.21, Section 3.5.2 5.5.8 CR occupancy factors Time (Hr) 9.1, Section 4.2.6 0-24 100 24-96 60 96-720 40 5.5.9 CR breathing rate 3.5E-04 m/sec 9.1, Section 4.2.6 5.5.10 CR detector specific 9.12, Table 13 efficiency Xe-133 4.1 1E7 cpm/jCi/cc Kr-85 2.51E8 cpni/iCi/cc 5.5.11 CR intake monitor 2480 cpm 9.6.14 alarm/trip setpoint Nuclear Common Revision 91

CALCULATION CONTINUA TION SHEET SHEET 17 of 34 CALC. NO.: S-C-ZZ-MDC-1951

REFERENCE:

LCRS03-05 I

Gopal Patel/NUCORE, ORIGINATOR, DATE J REV:

02/05/2004 0

Mark Drucker/NUCORE, REVIEWERIVERIFIER, DATE 02/05/2004 Design Input Parameter Value Assigned Reference 5.5.12 Unit I CR air intake x/Qs for Unit 1 MSSV Set 1 release Time X/Q (see/mr) 0-2 1.57E-02 9.5, Section 8.1 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 Unit 2 CR air intake X/Qs for Unit 1 MSSV Set I release Time X/Q (see/m3)_

0-2 6.98E-04 9.5, Section 8.1 2-8 5.66E-04 8-24 2.38E-04 24-96 1.65E-04 96-720 1.32E-04 5.6 Site Boundary Release Model Parameters 5.6.1 EAB atmospheric dispersion 1.3E-04 9.9, Table 5 factor (X/Q) (sec/m 3) 5.6.2 LPZ atmospheric dispersion factors (XIQs)

Time (Hr)

X/Q (sec/mr3) 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 (m'/sec) 3.5E-04 9.1, Section 4.1.3 5.6.4 LPZ breathing rates (m'/sec)

Time (Hr)

BR (m/.see) 9.1, Section 4.1.3 0-8 3.5E-04 8-24 1.8E-04---_-__

24-720 2.3E-04 Nuclear Common Revision 9

I CALCULATION CONTINUATION SHEET ISHEET 18 of 34 CALC. NO.: S-C-ZZ-MDC-1951

REFERENCE:

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

02/05/2004 0

l l

Mark Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02105/2004

6.0 CALCULATIONS

Due to the limitations of the RADTRAD code, the LRA 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 primary-to-secondary (P-T-S) leakage iodine activity release, the P-T-S leakage noble gas activity release, and the secondary liquid iodine release. The RADTRAD RFT file SGTRMSLBRFT.txt is interchangeably used for the 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 concentration RCS water density at cooled liquid conditions = 62.4 lb/ft3 (Ref. 9.1, Appendix G, Section 5.2)

RCS water mass = 12,446 ft3 x 62.4 lb/ft3 x 453.6 g/lbm = 3.523E+08 g SG Mass:

Reference 9.16, 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.

SG Mass For PC Iodine Activity Release:

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 lbs 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 =

SG Mass for Secondary Liquid Iodine Release:

When evaluating secondary liquid iodine releases, the larger Unit 2 SG mass is used to maximize the total iodine activity in the secondary liquid, which in turn yields higher secondary liquid release dose consequences.

For conservatism, the total Unit 2 SG liquid plus steam mass is treated as if it were only SG liquid:

Total SG volume = (4 SGs x 127,646 lb/SG) x (1/62.4) ft3 lb =

4 2' Total SG liquid mass = (4 SGs x 127,646 lb/SG) x 453.6 g/lb =

I Nuclear Common Revision 91l Nuclear Common Revision 9 I

CALCULATION CONTINUATION SHEET SHEET 19 of 34 CALC. NO.: S-C-ZZ-MDC-1951

REFERENCE:

LCR S03-05 l

~Gopal Patel/NUCORE,l ORIGINATOR, DATE REV:

02/05/2004 0

Mark Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/05/2004 6.2 Post-LRA 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 x (62.4 lb/ft3 x 0.0253 ft3/lb) x 1 ft3 = 0.188 cfm 7.482 gal 0-32 hours P-T-S leakage to four SGs =

6.2.2 SGs - Iodine Release Rates (P-T-S Leakage):

0-2 hours mass of steam released from four SGs to the environment = 6.55E+05 lb (Ref. 9.21, page 8)

Reactor pressure vessel (RPV) average temperature = 588.40F (Ref. 9.16, page 7), which is the 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.

Specific volume of steam = 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) 0-2 hours SG steaming rate = 6.55E+05 lb x 0.02313 ft3 /lb x (1/1.0) =

120 minutes 2-8 hours mass of steam released from SG to the environment = 5.40E+05 lb (Ref. 9.21, page 8) 2-8 hours SG steaming rate = 5.40E+05 lb x 0.02313 ft3/lb x (1/1.0) =

p m1 360 minutes 8-32 hours mass of steam released from SG = 2.40E+06 lb (Ref. 9.21, page 8) 8-32 hours SG steaming rate = 2.40E+06 lb x 0.02313 ft3/lb x (1/1.0) =

1440 minutes 0-2 hours iodine P-T-S leakage release from four SGs to the environment =

2-8 hours iodine P-T-S leakage release from four SGs to the environment =

8-32 hours iodine P-T-S leakage release from four SGs to the environment =

f s 1 UI 6.2.3 SGs Noble Gas Release Rates 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.

0-32 hours noble gas release rate to the environment = VN:

INuclear Common Revision9 Nuclear Common Revision 9 I

CALCULATION CONTINUATION SHEET SHEET 20 of 34 CALC. NO.: S-C-ZZMDC-1951

REFERENCE:

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

02/05/2004 0

Mark Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/05/2004 6.2.4 Secondary Liquid Iodine Release Rates:

The steaming rates for the secondary liquid iodine releases from the four steam generators are the same as those for the iodine release rates of P-T-S leakage (Section 6.2.2), which are then reduced by the iodine partition coefficient of 10 as follows:

0-2 hours secondary liquid iodine release from SGs to the environment = 126.25 x (1/10) =

2-8 hours secondary liquid iodine release from SGs to the environment = 34.70 x (1/10) =

i 8-32 hours secondary liquid iodine release from SGs to the environment = 38.55 x (1/10) =

6.3 CREACS Air Flow Rates:

Normal Flow Rate Notes "S" to the Reference 9.19 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 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 cfhm = 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 cfin) = 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 I0 percent uncertainty)

CREACS Recirculation Flow Rate With CR Monitors Preferentially Selecting Less Contaminated Air Intake CREACS ventilation flow rate = 8,000 cfm +/- 10% cefin (Ref. 9.6.7)

Minimum CREACS flow rate = 8,000 cfm - (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. 9.6.6) = 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. 9.6.9)

GL 99-02 (Ref. 9.15) 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% - ij)/safety factor = (100% - i)/2 Where 1 = charcoal filter efficiency to be credited in the analysis CREACS Charcoal Filter Nuclear Common Revision 9 INuclear Common Rtevision 91

CALCULATION CONTINUATION SHEET SHEET 21 of 34 CALC. NO.: S-C-ZZ-MDC-1951

REFERENCE:

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

02/05/2004 0

Mark Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/05/2004 2.5%= (100%-i)/2 5%= (100%--iT) i = 100% - 5% = 95%

HEPA Filter HEPA filter efficiency = 99% (Ref. 9.6.8). HEPA filter efficiency of 95% is used in the analysis Safety Grade Filter Efficiency Credited (%)

Filter Aerosol I

Elemental I

Organic CREACS 95 l

95 l

95 6.5 CR Intake Monitor Setpoint Evaluation:

CR Detector Specific Efficiencies:

Xe-133 = 4.11E7 cpm/jCi/cc (Ref. 9.12, Table 13)

Kr-85 = 2.51E8 cprn/pCi/cc (Ref. 9.12, Table 13)

The initial Xe-133 and Kr-85 concentrations at the CR intake for design basis LRA are calculated as follows:

Concentration = (RCS activity based on 5% assumed fuel failure)/(RCS volume) x (RCS leak rate) x (X/Q) x x (1.2% predicted fuel failure / 5% assumed fuel failure)

RCS Activity Xe-133 = 5.960E+05 Ci and Kr-85 = 8.438E+03 Ci (Table 4)

RCS Volume = 12,446 f3 (Ref. 9.6.2) and 0-2 hour X/Q = 1.57E-02 sec/m3 (Ref. 9.5, Section 8.1)

RCS Leak Rate of Noble Gas isotopes directly to the environment = 0.188 ft /min (Section 6.2.3)

Xe-133 Concentration at the CR intake

= (5.960E+05 Ci/12,446 ft3) x (0.188 t3/min) x (1.57E-02 sec/m3) x (1.2/5) x (1/60 min/sec)

= 5.65E-04 Ci/m3 = 5.65E-04 iCi/cc Kr-85 Concentration at the CR intake

= (8.438E+03 Ci/12,446 ft3) x (0.188 f 3/min) x (1.57E-02 sec/M3) x (1.2/5) x (1/60 min/sec)

= 8.OOE-06 Ci/m3 = 8.OOE-06,uCi/cc Count Rate Xe-133 = 5.65E-04 gCi/cc x 4.1 1E7 cpm/nuCi/cc = 23,221 cpm Count Rate Kr-85 = 8.00E-06 p.Ci/cc x 2.51E8 cpm/iCi/cc = 2,008 cpm Combined Count Rate of CR Intake Detector = 23,221 cpm + 2,008 cpm

= 25,229 cpm >> CR Monitor Setpoint of 2,480 cpm The Post-LRA concentration at the CR intake greatly exceeds the CR monitor setpoint value of 2,480 cpm, therefore the CR monitor will respond instantaneous.

I Nuclear Common Revision 91 Nuclear Common Revision 9 I

I CALCULATION CONTINUATION SHEET ISHEET 22 of 34 CALC. NO.: S-C-ZZ-MDC-1951

REFERENCE:

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

02/05/2004 0

Mark Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/05/2004 7.0 RESULTS

SUMMARY

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

RCP Locked Rotor Accident TEDE Dose (rem)

Receptor Location Control Room EAB LPZ P-T-S Iodine Release 1.29E+00 1.23E+00 5.01E-01 SLRAID00 (occurs @ 8.3 hrs)

SC Liquid Iodine Release 3.54E-03 6.24E-03 1.32E-03 SLRASLID00 (occurs @

0.0 hr)

Noble Gas Release 9.52E-03 2.24E-02 5.54E-03 SLRANG00 (occurs @ 0.0 hr)

Total 1.30E+00 1.26E+00 5.08E-01 Allowable TEDE Limit 5.OOE+00 2.50E+00 2.50E+00

8.0 CONCLUSION

S:

The Reactor Coolant Pump Locked Rotor accident results presented in Section 7.0 indicate that the EAB, LPZ, and CR doses due to a LRA are within their allowable limits.

I Nuclear Common Revision 91

C..

I CALCULATION CONTINUATION SHEET ISHEET 23 of 34 CALC. NO.: S-C-ZZ-MDC-1951

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

02/05/2004 0

l l

Mark Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/05/2004

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-I 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 3.7.1.4, Salem Unit 1/Unit 2 LCO for Plant System Activity

6.

Specification Surveillance Requirement 4.7.6.l.d.3, Salem Unit 1/Unit 2 CREACS Design Makeup Flow Rate

7.

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

8.

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

9.

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

10.

Specification 3.4.8, Salem Unit 1 LCO for Reactor Coolant System Specific Activity

11.

Specification 3.4.9, Salem Unit 2 LCO for Reactor Coolant System Specific Activity

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

14.

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

7.

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

8.

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

I CALCULATION CONTINUATION SHEET ISHEET 24 of 34 CALC. NO.: S-C-ZZ-MDC-1951

REFERENCE:

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

02/05/2004 0

l Mark Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/05/2004

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. 1EC-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.

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

13.

Westinghouse Calculation No. CN-CRA-93-149, Rev 0, Salem Locked Rotor Offsite Dose Analysis

14.

ASME Steam Tables, Sixth Edition

15.

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

16.

Nuclear Fuel Section Calculation File No. DS1.6-0453, Determination of Steam Release Flows for Input to the Radiological Dose Analysis

17.

Not Used.

18.

Letter FSE/SS-PSE-7561, Salem Units 1 & 2 Maximum RHRS "Cut-In" Time for Termination of Design Basis Event Steam Releases, 3/18/93

19.

SNGS Mechanical P&IDs:

a.

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

b.

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

20.

Letter FA-93-004, Dated 06/18/93, From A.J. Friedland to M.J. Fagan,

Subject:

LOF/LR for Salem Fuel Upgrade/Margin Recovery Program

21.

SNGS Calculation No. S-C-ZZ-MDC-1987, Rev. 1, Input Parameters for Salem AST Dose Calcs.

INuclear Common Revision 91 Nuclear Common Revision 9 I

I CALCULATION CONTINUATION SHEET ISHEET 25 of 34 CALC. NO.: S-C-ZZ-MDC-1951

REFERENCE:

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

02/05/2004 0

Mark Drucker/NUCORE, REVIEWERNVERIFIER, DATE 02/05/2004 10.0 TABLES:

Table I Iodine & Noble Gas Design Basis Core Inventory Core Design Basis Core Rated Core Isotope Inventory Thermal Inventory Power (Ci)

(MW,)

(Ci)

A B

Ax(Bxl.05)/3600 KR-83M 1.200E+07 3459 1.211E+07 KR-85 1.100E+06 3459 1.110E+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 2.000E+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 7.000E+05 3459 7.062E+05 XE-133 2.000E+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 4.000E+07 3459 4.036E+07 XE-138 1.600E+08 3459 1.614E+08 A From Reference 9.11, Table 2 at 3,600 MWt B From Reference 9.6.1 Nuclear Common Revision 9 I Nuclear Common Revision 91l

I CALCULATION CONTINUATION SHEET ISHEET 26 of 34 CALC. NO.: S-C-ZZ-MDC-1951

REFERENCE:

LCR S03-05 Gopal PateUJNUCORE, ORIGINATOR, DATE IREV:

02/05/2004 0

Mark Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/05/2004 Table 2 Post-LRA Design Basis Core Activity Release Design Basis Fuel Fuel Design Basis Core Gap Fraction Core Isotope Inventory Fraction Failed Activity Release (MW,)

(Ci)

A B

C AxBxC KR-83M 1.21 IE+07 0.05 0.05 3.027E+04 KR-85 1.110E+06 0.10 0.05 5.549E+03 KR-85M 2.623E+07 0.05 0.05 6.558E+04 KR-87 4.742E+07 0.05 0.05 1.185E+05 KR-88 6.759E+07 0.05 0.05 1.690E+05 1-131 9.988E+07 0.08 0.05 3.995E+05 1-132 1.412E+08 0.05 0.05 3.531E+05 I-133 2.018E+08 0.05 0.05 5.044E+05 1-134 2.220E+08 0.05 0.05 5.549E+05 1-135 1.917E+08 0.05 0.05 4.792E+05 XE-131M 7.062E+05 0.05 0.05 1.766E+03 XE-133 2.018E+08 0.05 0.05 5.044E+05 XE-133M 2.926E+07 0.05 0.05 7.314E+04 XE-135 5.044E+07 0.05 0.05 1.261E+05 XE-135M 4.036E+07 0.05 0.05 1.009E+05 XE-138 1.614E+08 0.05 0.05 4.036E+05 A From Table I B From Reference 9.1, Table 3 C From Design Input 5.3.5 I Nuclear Common Revision 9

I CALCULATION CONTINUATION SHEET ISHEET 27 of 34 CALC. NO.: S-C-ZZ-MDC-1951

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LCR S03-05 Gopal Patel/NUCORE,<

ORIGINATOR, DATE REV:

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Mark Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/05/2004 Table 3 Total 1% Failed Fuel Primary Coolant Activity in RCS 1% Fuel PC PC Defects Mass In Total Isotope PC Activity RCS Activity Concentration (Pci/g)

(g)

(Ci)

A B

C=AxB/1E+6 KR-83M 4.OOOE-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 1-131 2.80E+00 3.523E+08 9.864E1+02 I-132 2.80E+00 3.523E+08 9.864E+02 1-133 4.20E+00 3.523E+08 1.480E+03 I-134 5.70E-01 3.523E+08 2.008E+02 I-135 2.30E+00 3.523E+08 8.103E+02 XE-13 IM 2.1003E+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

.r Nuclear Common Revision 9 INclear Common Revision 91l

a-I CALCULATION CONTINUATION SHEET ISHEET 28 of 34 CALC. NO.: S-C-ZZ-MDC-1951

REFERENCE:

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

02/05/2004 0

Mark Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/05/2004 Table 4 Total Post-LRA Activity Released To RCS PC Design Basis Total RADTRAD Total Core Post-LRA Nuclide Isotope Activity Activity Activity Inventory Release Released To File RCS (Ci)

(Ci)

(Ci)

(Ci)

A B

C=A+B C

KR-83M 1.409E+02 3.027E+04 3.041E+04

.3041E+05 KR-85 2.889E+03 5.549E+03 8.438E+03

.8438E+04 KR-85M 5.989E+02 6.558E+04 6.618E+04

.6618E+05 KR-87 3.523E+02 1.185E+05 1.189E+05

.1189E+06 KR-88 1.057E+03 1.690E+05 1.700E+05

.1700E+06 I-131 9.864E+02 3.995E+05 4.005E+05

.4005E+06 I-132 9.864E+02 3.531E+05 3.541lE+05

.3541E+06 1-133 1.480E+03 5.044E+05 5.059E+05

.5059E+06 1-134 2.008E+02 5.549E+05 5.551lE+05

.5551E+06 I-135 8.103E+02 4.792E+05 4.800E+05

.4800E+06

-XE-131M 7.398E+02 1.766E+03 2.505E+03

.2505E+04 XE-133 9.160E+04 5.044E+05 5.960E+05

.5960E+06 XE-133M 5.989E+03 7.314E+04 7.913E+04

.7913E+05 XE-135 2.995E+03 1.261E+05 1.291E+05

.1291E+06 XE-135M 1.726E+02 1.009E+05 1.01 IE+05

.1011E+06 XE-138 2.149E+02 4.036E+05 4.038E+05

.4038E+06 A From Table 3 B From Table 2 I Nuclear Common Revision 91 Nuclear Common Revision 9 I

I.

I CALCULATION CONTINUATION SHEET ISHEET 29 of 34 CALC. NO.: S-C-ZZ-MDC-1951

REFERENCE:

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

02/05/2004 0

Mark Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/05/2004 Table 5 Iodine Isotopic Dose Conversion Factors Isotopic Conversion Iodine Dose Factor Dose Isotope Conversion Conversion Factor Factor (Sv/Bq)

(rem/Ci/Sv/Bq)

(rem/Ci)

A B

C=AxB I-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 I-134 2.88E-10 3.70E+12 1.07E+03 1-135 8.46E-09 3.70E+12

.3.13E+04 A From Reference 9.7, Page 136 Table 6 Iodine Scaling Factors Pre-accident Iodine Spike & Equilibrium Iodine Concentration 1% Failed Fuel Iodine Iodine Dose Product Isotope Activity Conversion Concentration Factor gCi.rem/Ci.g (WCi/g)

(rem/Ci)

(rem)

A B

(AxB) 1-131 2.80E+00 1.08E+06 3.02E+06 1-132 2.80E+00 6.44E+03 1.80E+04 1-133 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 9.11, Table 4; and B from Table 5

. 131 DE Based on 1% FF Iodine Concentration l

3.58E+00 Iodine Scaling Factor Based on 1.0 pCi/g DE 1-131 2.791E-01 INuclear Common Revision 91 Nuclear Common Revision 9 I

A I

CALCULATION CONTINUATION SHEET SHEET 30 of 34 CALC. NO.: S-C-ZZ-MDC-1951

REFERENCE:

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

02/05/2004 0

Mark Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/05/2004 Table 7 RCS Iodine Concentration Based on 1.0 gtCi/g DE 1-131 1% Failed Fuel Iodine 1.0 iCi/g Iodine Scaling DE 1-131 Isotope Activity Factor Activity Concentration 1.0 pCi/g Concentration (ACi/g)

DE 1-131 (110/g)

A B

C=AxB 1-131 2.80E+00 2.791E-01 7.814E-01 1-132 2.80E+00 2.791E-01 7.814E-01 1-133 4.20E+00 2.79 1E-01 1.172E+00 1-134 5.70E-01 2.791E-01 1.591E-01 1-135 2.30E+00 2.791E-01 6.419E-0 1 A From Reference 9.11, Table 4 B Scaling Factor Based on 1.0 pCi/g DE 1-131 From Table 6 Table 8 Secondary Side Iodine Activity 1.0 PiCig Tech Spec Coolant Mass Secondary RADTRAD DE I-131 Secondary In Steam Side Nuclide Isotope Activity Coolant Generator Activity Inventory Concentration Activity Limit File (1C0/g)

(XiCi/g)

(g)

(Ci)

(Ci)

A B

C AxBxC/1E6 1-131 7.814E-01 0.10 2.316E+08 18.098

.181OE+02 1-132 7.814E-01 0.10 2.316E+08 18.098

.181 OE+02 1-133 1.172E+00 0.10 2.316E+08 27.147

.2715E+02 1-134 1.591E-01 0.10 2.316E+08 3.684

.3684E+01 1-135 6.419E-01 0.10 2.316E+08 14.866

.1487E+02 A From Table 7 B From Reference 9.6.5 C From Section 6.1 I Nuclear Common Revision 91 Nuclear Common Revision 9 I

I CALCULATION CONTINUATION SHEET ISHEET 31 of 34 CALC. NO.: S-C-ZZ-MDC-1951

REFERENCE:

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

02/05/2004 0

Mark Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/05/2004 11.0 FIGURES:

Figure 1: RADTRAD Nodalization For Post-LRA P-T-S Iodine & Noble Gas Releases P-T-S Leakage Iodine Release

. _ P-T-S Leakage Noble Gas Release l Nuclear Common Revision 9

S.

CALCULATION CONTINUATION SHEET ISHEET 32 of 34 CALC. NO.: S-C-ZZ-MDC-1951

REFERENCE:

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

02105t2004 0

Mark Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/05/2004 Figure 2: RADTRAD Nodalization For Post-LRA Secondary Liquid Iodine Release INuclear Common Revision 91l Nuclear Common Revision 9 I

I CALCULATION CONTINUA TION SHEET SE33 4

CALC. NO.: S-C-ZZ-MDC-1951

REFERENCE:

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

02/05/2004 0

Mark Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/05/2004 Figure 3 - RADTRAD Nodalization For CR Response With Monitors Preferentially Selecting Less Contaminated Air Intake Nuclear Common Revision 9 I INuclear Common Revision 9

l CALCULATION CONTINUATION SHEET SHEET 34 of 34 CALC. NO.: S-C-ZZ-MDC-1951

REFERENCE:

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

02/05/2004 0

Mark Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/05t2004 12.0 AFFECTED DOCUMENTS:

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

Documents to be voided:

Vendor Technical Document No. 321041, Rev. 2, Radiological Dose Consequence At Control Room For A RCP Locked Rotor Accident, With Original CR Ventilation Design.

Vendor Technical Document No. 321042, Sheet 1, Rev. 4, Radiological Dose Consequence At EAB/LPZ, And Control Room For A RCP Locked Rotor Accident, With Modified CR Ventilation Design Vendor Technical Document No. 321042, Sheet 2, Rev. 2, Radiological Dose Consequence At EAB/LPZ, And Control Room For A RCP Locked Rotor Accident, With Modified CR Ventilation Design Vendor Technical Document No. 323179, Rev. 1, Thirty (30) Day Control Room Dose Following A RCP Locked Rotor Accident With Unfavorable CR Intake/Meteorology And Both EACS Units Operational Documents to be revised:

UFSAR Section 15.4.5, Single Reactor Coolant Pump Locked Rotor and Reactor Coolant Pump Shaft Break 13.0 ATTACHMENTS:

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

Calculation No: S-C-ZZ-MDC-1951, OIR1.

Comment Resolution Form 2 - Mark Drucker Certification for Design Verification Form-i RCPD Form-I Nuclear Common Revision 9 INuclear Common Revision 91l

Attachment A S-C-ZZ-MDC-1951, Rev. 0 1 Diskette With Various Electronic Files