ML042740514

Eab, LPZ, & CR Doses - Steam Generator Tube Rupture (SGTR) Accident - AST

ML042740514
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-1949, Rev OIR2
Download: ML042740514 (42)

Text

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

CALC NO.: S-C-ZZ-MDC-1949 CALCULATION COVER SHEET Page 1 of 40 REVISION: 01R2 CALC. TITLE:

I EAB, LPZ, & CR Doses - Steam Generator Tube Rupture (SGTR) Accident - AST

1. SHTS (CAL) 40 ATT / # SHTS:

111

1. IDVl50.59 SHTS: I 21-I #TOTAL SHTS:

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CHECK ONE:

go/34P E FINAL 0 INTERIM (Proposed Plant Change) l FINAL (Future Confirmation Req'd) rl VOID SALEM OR HOPE CREEK:

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

[-OQ ElQs EQsh OF OR 0 STATION PROCEDURES IMPACTED, IF SO CONTACT RELIABILITY ENGINEER a CDs INCORPORATED (IF ANY):

DESCRIPTION OF CALCULATION REVISION (IF APPLY):

N/k 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 Steam Generator Tube Rupture (SGTR) 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 SGTR accident are within their allowable TEDE dose limits.

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

OIR1 Editorial comments incorporated.

OIR2 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 15 6.0 Calculations 20 7.0 Results Summary 26 8.0 Conclusions

-28 9.0 References 28 10.0 Tables 30 11.0 Figures 37 12.0 Affected Documents 40 13.0 Attachments 40 INuclear Commo Revision 9 l Nuclear Common Revision 9

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Mark Drucker/NUCORE, REVIEWERIVERIFIER, DATE 02/05/2004

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 Steam Generator Tube Rupture (SGTR) 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 SGTR accident 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 SGTR accident is analyzed using plant specific design and licensing bases inputs, which are compatible to the TEDE dose criteria. The SGTR analysis is performed using the guidance in Regulatory Guide 1.183 and its Appendix F (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 SGTR accident. The RADTRAD code is documented in NUREG/CR-6604 (Ref. 9.2). The RADTRAD code is maintained as Software ID Number A-O-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 provides emergency filtration and air conditioning services to the combined CR for the Salem 1 & 2 plants (Ref. 9.19). Each CREACS train is safety related and required to be operable by a Technical Specification (Refs. 9.6.14 and 9.6.15). 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 normal intake exceeds the setpoint value (see Section 6.7 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. Per Section 4.0, this calculation addresses the reactor coolant activity concentrations corresponding to (l) a preaccident iodine spike and (2) a concurrent iodine spike. There is no potential fuel failure during this event.

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Mark Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/05/2004 3.1 Preaccident Iodine Spike Release In the preaccident iodine spike release scenario, a reactor transient has occurred prior to the postulated SGTR accident and has raised the primary coolant (PC) iodine concentration to the maximum value permitted by technical specifications. The PC iodine concentration of 60 iLCi/g Dose Equivalent (DE) of I-131 is obtained from Figure 3.4-1 of technical specifications (Ref. 9.6.7). The plant-specific PC iodine concentration corresponding to 1% fuel defects is obtained from Reference 9.11 (Table 4), and listed in Table 2. The iodine dose conversion factors are developed in Table I using the information in Reference 9.7 and used in Table 2 to establish the iodine scaling factor based on the iodine concentration of I pCi/g DE I-13 1, which is used to convert the iodine concentration of 1% fuel defects to 1 iLCi/g DE I-13 1 in Table 3. The total isotopic iodine activity in the RCS is conservatively calculated in Table 4 using the PC mass determined at "cooled" liquid conditions (section 6.1). Since the Salem 1 & 2 PC normal noble gas isotopic concentrations are not readily available, the 1 OO/E-Bar noble gas concentrations are not calculated. The PC noble gas concentrations corresponding to 1% fuel defects are conservatively used in Table 5 to determine the total RCS noble gas activity. The total isotopic activities released to the RCS are calculated in Table 6, which are used to develop the RADTRAD Nuclide Inventory File (NIF) SGTRPREdef.txt for the preaccident iodine spike case. The post-SGTR activity transport and CR response models are shown in Figures 1 & 3, which are used to determine the post-SGTR radiological consequences using the Salem plant-specific as-built design inputs and assumptions listed in Sections 4 & 5. The CR response model utilizes the Salem CR air intake monitor to preferentially select the less contaminated air intake after I minute (Ref. 9.10). A delay time of I minute for CR pressurization mode to be fully operational is conservatively assumed in this analysis. The EAB, LPZ, and CR doses are summarized in Section 7.1 for the preaccident iodine spike case. The following paragraphs describe the activity transport paths from the one faulted and three intact steam generators (SGs). These activity transport paths are applicable to both the preaccident and concurrent iodine spike cases.

Faulted SG Release The PC is postulated to be released into the faulted SG having a single SG tube ruptured. The resultant equilibrium break flow is assumed to continue for 30 minutes after the initiation of the event, at which time it is assumed that the operator isolates the faulted S/G and depressurizes the RCS to the faulted S/G pressure ending the transfer of reactor coolant to the secondary side of the faulted S/G and the release to the atmosphere. An assessment was done (Ref 6.25) that determined that with a more sophisticated LOFTTR2 analysis the break flow termination could be extended to 55 minutes without exceeding the integrated break flow assumed in the original analysis (Ref 6.16). To maximize the offsite doses it is assumed that offsite power is lost so that the main steam condensers are not available. Iodine activity is released from the faulted SG to the environment in proportion to the steaming rate. The iodine activity from the P-T-S leakage is assumed to be directly released in the SG volume and partitioning of iodine is not credited. PC noble gas activity entering the faulted SG is assumed to be released to the environment without holdup or decontamination in the secondary system. For the secondary liquid iodine release from the faulted SG, 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 F, Section 5.6 and Appendix E, Section 5.5.4) to maximize the resulting dose and the SG steaming rates are reduced accordingly for the secondary liquid iodine release.

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CORE, REVIEWER/VERIFIER, DATE 02/05/2004 Intact SGs Release Due to the SGTR transient the reactor shuts down and the RCS loses a substantial amount of coolant inventory with the potential of over flooding the faulted SG. In order to remove decay heat the plant begins to release the secondary coolant via the intact SGs' atmospheric relief valves. The steam release from the three intact SGs continues for 32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br /> until the Residual Heat Removal (RHR) system is aligned to dissipate heat (Ref. 9.18).

During the 32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br /> before the RHR system is initialized, the PC is assumed to leak into the three intact SGs at the maximum technical specification leak rate of 1 gpm (Refs. 9.6.4 & 9.6.5). The PC noble gases are released directly to the environment from the RCS without holdup while the iodine activity is directly released from intact SG in proportion to the steaming rate. For the secondary liquid iodine release from the three intact 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 F, Section 5.6 and Appendix E, Section 5.5.4) to maximize the resulting dose and the SG steaming rates are reduced accordingly.

3.2 Concurrent Iodine Spike:

In the concurrent iodine spike release scenario, the primary system transient associated with the SGTR causes an iodine spike in the primary system. The increase in primary coolant iodine concentration is estimated using a spiking model that assumes that the iodine release rate from the fuel rods to the primary coolant (expressed in curies per unit time) increases to a value 335 times greater than the release rate corresponding to the iodine concentration at the equilibrium value (typically 1.0 piCi/gm DE I-13 1) specified in technical specifications (i.e., concurrent iodine spike case) (Ref. 9.1, Appendix F, Section 2.2). The isotopic iodine appearance rates are calculated in Section 6.2 and Table 7 using the specific activity production and removal rates based on the most conservative letdown system parameters. The total isotopic iodine activity is calculated in Table 8 using the spike duration of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. The total isotopic iodine and noble gas activities are listed in table 9 and used to develop the RADTRAD NIF SGTRCONdeftxt for the concurrent iodine spike case. The post-SGTR activity transport and CR response models are shown in Figures 2 & 3, which are used to determine the post-SGTR radiological consequences using the Salem plant-specific as-built design inputs and assumptions listed in Sections 4 & 5. The CR response model utilizes the Salem CR air intake monitor to preferentially select the less contaminated air intake after 1 minute (Ref. 9.10). The EAB, LPZ, and CR doses are summarized in Section 7.2 for the concurrent iodine case. The activity in the fuel is postulated such that the 99% of all fuel activity is released into the RCS within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> at a calculated rate of 9.60 cfm (Section 6.4). The activity transport from the RCS to the environment and the CR response are the same as that for the preaccident iodine spike case.

3.3 Steam Generator Iodine Activity Release:

The iodine activity in the SG is calculated based on a secondary side liquid activity concentration of 0.1 [tCi/g DE I-131 (Ref. 9.6.6) using a larger SG volume. The iodine partition coefficient of 10 is conservatively used in lieu of the value of 100 recommended in Regulatory Guide 1.183 (Ref. 9.1, Appendix F, Section 5.6 and Appendix E, Section 5.5.4) to maximize the resulting dose. The iodine activity is assumed to be released to the environment at the SG steaming rates.

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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 SGTR accident. The following sections address the applicability of these modeling assumptions to an SNGS Units 1 and 2 SGTR 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 Terms There is no fuel damage postulated for the SGTR accident (Ref. 9.17, page 4). Per RG 1.183 (Ref. 9.1, Appendix F, Section 2) since no or minimal fuel damage is postulated for the limiting event, the released activity is the maximum coolant activity allowed by the technical specifications. Two cases of iodine spiking, a preaccident case and a concurrent case, are assumed.

4.1.1 Preaccident Iodine Spike Source Term Consistent with RG 1.183 (Ref. 9.1, Appendix F, Section 2.1) the first iodine spiking case assumes that a reactor transient has occurred prior to the postulated SGTR and has raised the primary coolant iodine concentration to the maximum value permitted by the technical specifications.

Per Technical Specification (TS) LCO for Specific Activity (Refs. 9.6.1 & 9.6.2), the maximum primary coolant iodine concentration permitted by the technical specifications at full power is shown in TS Figure 3.4-1 to be 60 piCi/gm Dose Equivalent (DE) I-131. Technical Specification 1.10 (Ref. 9.6.13) defines DE I-131 as that concentration of 1-131 (in [tCi/g) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, 1-133, 1-134, and I-135 actually present using the thyroid dose conversion factors specified in Table III of TID-14844; this analysis calculates DE-131 in terms of the thyroid dose conversion factors specified in Federal Guidance Report 12 (Ref. 9.8). The preaccident iodine spike isotopic iodine concentrations are calculated in Table 4.

4.1.2 Concurrent Iodine Spike Source Term Consistent with RG 1.183 (Ref. 9.1, Appendix F, Section 2.2) the second iodine spiking case assumes that the primary system transient associated with the SGTR causes an iodine spike in the primary system. The increase in primary coolant iodine concentration is estimated using a spiking model that assumes that the iodine release rate from the fuel rods to the primary coolant (expressed in curies per unit time) increases to a value 335 times greater than the release rate corresponding to the iodine concentration at the equilibrium value specified in technical specifications (i.e., concurrent iodine spike case). The assumed iodine spike duration is 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

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Mark Drucker/NUCORE, REVIEWER/NIERIFIER, DATE 02/05/2004 Per Technical Specification LCO for Specific Activity (Refs. 9.6.1 & 9.6.2), the equilibrium primary coolant iodine concentration permitted by the technical specifications is 1.0 liCi/gm DE I-13 1. The concurrent iodine spike isotopic iodine activity release rates are calculated in Table 7.

4.1.3 Noble Gas Source Term In both preaccident and concurrent iodine spike cases the primary coolant noble gas (Xenon and Krypton) concentrations are assumed to correspond to the maximum coolant activity allowed by the technical specifications.

Per Technical Specification LCO Specific Activity (Ref. 9.6.1 & 9.6.2), the equilibrium primary coolant non-iodine concentration permitted by the technical specifications is Il0/E-BAR. The parameter E-BAR is defined as the average, weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling, of the sum of the average beta and gamma energies per disintegration, in MeV, for isotopes, with half lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant.

Table 2 calculates a 1 percent failed fuel primary coolant iodine concentration of 3.58 iiCi/gm DE I-13 1, which is conservatively greater than the TS limit of 1.0 [tCi/gm DE I-13 1. Since 1 percent failed fuel introduces more iodine activity into the primary coolant than is allowed by the Tech Specs, it can be conservatively assumed that 1 percent failed fuel introduces more non-iodine activity into the RCS than is allowed by the Tech Spec limit of 1 00/E-bar. Therefore, the PC noble gas concentrations corresponding to 1% fuel defects are conservatively used in Table 5 to determine the total RCS noble gas activity.

4.1.4 Activity Release from the Fuel into the Primary Coolant Consistent with RG 1.183 (Ref. 9. 1, Appendix F, Section 3) the activity released from the fuel (for both the preaccident and concurrent iodine spike cases) is assumed to be released instantaneously and homogeneously through the primary coolant.

4.1.5 Iodine Chemical Form Consistent with RG 1.183 (Ref. 9.1, Appendix F, 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 released during normal operations, including iodine spiking.

4.2 Transport 4.2.1 Primary-to-Secondary (P-T-S) Leak Rates Consistent with RG 1.183 (Ref. 9.1, Appendix F, Section 5.1) the primary-to-secondary leak rate is apportioned between affected and unaffected steam generators in such a manner that the calculated dose is maximized.

Technical Specification LCO for RCS Operational Leakage (Refs. 9.6.4 & 9.6.5) specifies a maximum limit of one gpm total primary-to-secondary leakage through all steam generators and 500 gallons per day through any one steam generator. The total P-T-S leak rate of I gpm into the three intact SGs is postulated to maximize the I Nuclear Common Revision 9 1 I Nuclear Common Revision 9

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Mark Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/05/2004 dose because the P-T-S leakage continues for a longer duration (32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br />) in the intact SGs in comparison to the faulted SG (30 minutes).

4.2.2 Primary-to-Secondary Leak Primary Coolant Density Consistent with RG 1.183 (Ref. 9.1, Appendix F, 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., lbm/hr) is assumed to be 1.0 gm/cc (62.4 Ibrm/fl3). Conversely, the temperature dependent specific volumes are used for converting the mass flow rates into volumetric flow rates to maximize the releases and consequently the doses.

4.2.3 Primary-to-Secondary Leak Duration Consistent with RG 1.183 (Ref. 9.1, Appendix F, 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 0F. In this analysis primary coolant is conservatively assumed to leak into the intact 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 F, Section 5.4) the release of fission products from the secondary system should be evaluated with the assumption of 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 All noble gases radionuclides released from the primary system are assumed to be released to environment without reduction or mitigation.

4.2.6 Iodine & Particulate Transport Model Consistent with RG 1.183 (Ref. 9.1, Appendix F, Section 5.5) the transport model described in Regulatory Positions 5.5 and 5.6 of Appendix E should be utilized for iodine and particulates. The post-SGTR thermal hydraulic condition in the faulted SG is such that a large amount of PC released through the ruptured tube excessively increases the coolant mass inventory, which eliminates the possibility of SG dryout condition and subsequently flashing of PC in the faulted SG. The coolant mass in the intact SGs provides adequate tube submergence to eliminate the flashing of PC in the primary-to-secondary leakage. Therefore, the transport model described in Regulatory Positions 5.5 & 5.6 of Appendix E is not applicable to the post-SGTR thermal-hydraulic conditions exist in the faulted and 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 Nuclear Common Revision 9

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Mark Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/05/2004 the low population zone (LPZ). The following sections address the applicability of this guidance to the SNGS Units 1 and 2 SGTR accident 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 5.3.11 & 5.3.12.

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).

4.3.3 Offsitc 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

-- eonversion 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 CFR50.67 (Ref. 9.3). For the SGTR accident the postulated EAB doses should not exceed the criteria established in RG 1.183 Table 6:

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Mark Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/05/2004 EAB Dose Acceptance Criterion (preaccident iodine spike case):

25 Rem TEDE EAB Dose Acceptance Criterion (concurrent iodine spike case):

2.5 Rem TEDE Per RG 1.183 Table 6, the SGTR 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 SGTR accident the postulated LPZ doses should not exceed the criteria established in RG 1.183 Table 6:

LPZ Dose Acceptance Criterion (preaccident iodine spike case):

25 Rem TEDE LPZ Dose Acceptance Criterion (concurrent iodine spike case):

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

4.3.7 Effluent Plumc 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 I and 2 SGTR accident 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, I Nuclear Common Revision 9 l

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• 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 SGTR accident 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 SGTR accident are insignificant compared to those due to a LOCA (see the core release fractions for LOCA and non-LOCA design basis accidents in Tables 1 and 3 of Reference 9.1). Therefore, these direct dose contributions are considered to be insignificant and are not evaluated for a SGTR accident.

4.4.2 Control Room Source Term Consistent with RG 1.1 83 (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 nonconservative 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).

4.4.4 Control Room Response Consistent with RG 1.183 (Ref. 9. 1, Section 4.2.4), credit for engineered TaTe1fe-tau'r (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 6.7 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.

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Mark DruckerfNUCORE, REVIEWER/VERIFIER, DATE 02/05/2004 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 I 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 4.2.7), the control room doses are calculated using the offsite dose analysis dose conversion factors identified in RG 1.183 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, DDEO, to a finite cloud dose, DDErinite, where the control room is modeled as a hemisphere that has a volume, V, in cubic feet, equivalent to that of the control room:

DDEfinit, = (DDE. x V0-338) / 1173 4.4.8 Control Room Operator Dose Acceptance Criteria Consistent with RG 1.183 (Ref. 9.1, Section 4.4), for the SGTR accident the postulated CR doses should not exceed the 5 Rem TEDE criterion established in 10 CFR 50.67 (Ref. 9.3):

CR Dose Acceptance Criterion (preaccident iodine spike case):

5 Rem TEDE CR Dose Acceptance Criterion (concurrent iodine spike case):

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

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Mark Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/05/2004 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 %/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 F 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, REVIEWER/VERIFIER, DATE 02/05/2004 Design Input Parameter Value Assigned Reference Steam Generator Tube Rupture Accident Parameters 5.3 Source Terms 5.3.1 Licensed reactor core power 3,459 MWt 9.6.8 level 5.3.2 Primary Coolant (PC) 1.0 jiCi/gm DE Iodine-131 9.6.1 & 9.6.2 (iodine) specific activity in the Technical Specifications 5.3.3 PC (non-iodine) specific 100 / E-bar 9.6.1 & 9.6.2 activity in the Technical Specifications 5.3.4 Maximum PC (iodine) 60 p.Ci/gm DE Iodine-131 at 9.6.7 specific activity permitted by the 100 percent rated thermal power Technical Specifications 5.3.5 Concurrent Iodine Spiking 335 (for an SGTR accident) 9.1, Appendix F, Section 2.2 Factor 5.3.6 Duration of concurrent 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> (for an SGTR accident) 9.1, Appendix F, Section 2.2 iodine spike 5.3.7 Secondary coolant (iodine) 0.1 jiCi/gm DE Iodine-131 9.6.6 specific activity in the Technical Specifications 5.3.8 Nominal reactor coolant 12,446 ft3 9.6.3 system (RCS) volume 5.3.9 Total primary-to-secondary 1 gpm 9.6.4 & 9.6.5 leakage through all steam generators (SGs) 5.3.10 Total primary-to-secondary 500 gallons/day 9.6.4 & 9.6.5 leakage through any one SG 5.3.11 PC iodine activity concentration based on 1% fuel defects 9.11, Table 4 Isotope Activity (p.Ci/g)

Isotope Activity (giCilg)

Isotope Activity (jiCi/g)

I-131

-2t8-I-133 4.2 1-135 2.3 1-132 2.8 1-134 0.57 5.3.12 PC noble gas activity based on 1% fuel defects 9.11, Table 4 Isotope Activity (gCi/g)

Isotope Activity (gCi/g)

Isotope Activity (jiCi/g)

Kr-83M 4.OE-01 Kr-88 3.OE+00 Xe-135M 4.9E-01 Kr-85M 1.7E+00 Xe-131M 2.1E+00 Xe-135 8.5E+00 Kr-85 8.2E+00 Xe-133M 1.7E+01 Xe-138 6.1E-01 Kr-87 l.OE+00 Xe-133 2.6E+02 Nuclear Common Revision 9

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Mark Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/05/2004 Design Input Parameter Value Assigned Reference 5.3.13 Steam Generator Dilution 119,233 Ibm per Ul Model F SG 9.16, pages 7 and 8 Mass 127,646 ibm per U2 Model 51 SG 5.4 Activity Transport Models (Figures 1 & 2) 5.4.1 PC released from RCS to 137,246.2 lb 9.17, page 8 faulted SG (i.e., rupture flow) 5.4.2 Steam mass released from 9.24, page 8 faulted SG to the environment 0.0 - 0.5 hr 56,500 lb 5.4.3 PC leakage to intact SGs 1 gpm 9.6.4 & 9.6.5 0.188 cfm Section 6.3 5.4.4 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, page 4 duration 5.4.5 Steam mass released from 9.24, page 8 intact SGs to the environment 0 - 2 hr 465,000 lbs 2.- 8 hr 1,055,000 lbs 8 - 24 hr 1,503,000 lbs 24 - 30 hr 477,000 lbs 30 - 32.0 hr 150,333 lbs=

[(451,000) x (32 -30)/(36 -30)]

5.4.6 SG liquid iodine partition 100 9.1, Appendix F, Section 5.6 and coefficient 10 (used in the analysis)

Appendix E, Section 5.5.4 5.5 Control Room Model Parameters (Figure 3) 5.5.1 CR volume 81,420 ftW 9.10, page 33 5.5.2 CR normal flow rate 1,320 cfm (for two air intakes)

Section 6.5 5.5.3 CREACS makeup flow rate 2,200 cfm 9.6.9 5.5.4 CREACS recirc flow rate 8,000 cfm +/- 10% cfm 9.6.10 with one train operating 5,000 cfm (used in analysis)

Section 6.5 5.5.5 CREACS charcoal filter 95%

Section 6.6 efficiency l

5.5.6 Delay time for CR 1 minute Assumed pressurization 5.5.7 CREACS HEPA filter 99%

9.6.11 Section 6.6 efficiency 95% (used in analysis) 5.5.8 CR unfiltered inleakage 150 cfm 9.24, Section 3.5.2 Nuclear Common Revision 9

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Mark Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/05/2004 Design Input Parameter Value Assigned l

Reference 5.5.9 CR occupancy factors Time (Hr)

__9.1, Section 4.2.6 0-24 100 24-96 60 96-720 40 5.5.10 CR breathing rate 3.5E-04 m'/sec 9.1, Section 4.2.6 5.5.11 Unit 1 CR air intake x/Qs for Unit I MSSV Set 1 release Time X/Q (sec/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 5.5.12 Unit 2 CR air intake X/Qs for Unit I MSSV Set 1 release Time XJQ (sec/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 5.5.13 CR allowable dose limit 5 rem TEDE 9.3 5.5.14 CR detector specific 9.12, Table 13 efficiency Xe-133 4.11 E7 cpm/,lCi/cc Kr-85 2.51E8 cpm/liCi/cc 5.5.15 CR intake monitor 2480 cpm 9.6.16 alarm/trip setpoint 5.6 Site Boundary Release Model Parameters 5.6.1 EAB atmospheric dispersion 1.30E-04 factor (X/Q) (sec/m3 )

1 5.6.2 LPZ atmospheric dispersion factors (X/Qs)

T 9.9, Table 5

_z Time (Hr)

X/Q (sec/m3) 9.9, Table 5 0-2 1.86E-05 2-8 7.76E-06 8-24 5.01E-06 24-96 11.94E-06 96-720 4.96E-07 I

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Mark Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/05/2004 Design Input Parameter Value Assigned Reference 5.6.3 EAB breathing rate (mP/sec) 3.5E-04 9.1, Section 4.1.3 5.6.4 LPZ breathing rates (m3/sec)

Time (Hr)

BR (m3/sec) 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 l

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6.0 CALCULATIONS

This calculation addresses two cases, a preaccident iodine spiking case and a concurrent iodine spiking case.

Due to the limitations of the RADTRAD code, the SGTR dose consequences for each case are determined by summing together the results from sets of three RADTRAD code analyses that separately calculate the dose contributions from the primary-to-secondary (P-T-S) leakage preaccident or concurrent iodine spike 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 iodine release by setting the iodine release fraction to 1.0 and for 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 @ T,,e of 5730F (Ref. 9.6.3)

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 F, Section 5.2)

RCS water specific volume at 5730F = 0.02253 ft3/lb (Ref. 9.14, page 182)

RCS water density at 5730F = 1/0.02253 = 44.39 lb/ft RCS water mass = 12,446 ft3 x 62.4 lb/ft3 x 453.6 g/lbm = 3.523E+08 g used for calculating the RCS activity in Table 4 RCS water mass = 12,446 ft3 x 44.39 lb/ft3 x 453.6 g/lbm = 2.506E+08 g used for calculating the iodine appearance rate in Section 6.2 SG Mass:

Reference 9.16, pages 7 and 8, indicates that Salem Unit I has Model F steam generators each having a total liquid plus steam mass of 119,233 Ibs, and Salem 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.

Faulted SG volume = 119,233 lb x (1/62.4) ft3/lb =k-Intact SGs total volume = 3 x 1,911 ft3 =iEi INuclear Common Revision 9 Nuclear Common Revision 9

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l Mark Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/05/2004 SG Mass For Secondary Liquid Iodine Activity 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 Faulted + Intact SGs volume = (4 SGs x 127,646 lb/SG) x (1/62.4) ft3 b =/l Total Faulted + Intact SGs liquid mass = (4 SGs x 127,646 lb/SG) x 453.6 glib = 3 6.2 Iodine Spike Appearance Rate:

Letdown coolant temperature before entering demineralizer = 130 0F (Ref. 9.13, page 4)

Letdown demineralizer decontamination factor (DF) = 10 (Ref. 9.13, page 4)

Specific volume of letdown coolant at 130 0F = 0.016247 ft /lb (Ref. 9.14, page 185)

RCS mass = 3.523E+08 g (Section 6.1)

Letdown flow rate = 165 gpm (Ref. 9.15)

=165 gal x 1 ft3 x

1 lb x

453.6 1

x I min

=10,263 g/sec min 7.481 gal 0.016247 ft3 lb 60 sec Specific Activity = Production Rate /Removal Rate (Ref. 9.13, page 4)

Xtotal = Xpurification + Xdecay = [(letdown flow / RCS mass) (1 - 1/DF)] + [decay removal rate]

Letdown purification rate = Xpurification = Letdown flow rate/RCS mass (1 - 1/DF)

=

10,263 g x

1 x (1 - 1/10) = 3.686E-05 (sec)'

sec 2.506E+08 g Iodine Appearance Rate = Ai x Xtotal = Ai (3.686E-05 see' + 2-decay sec-)

Where Ai = total isotopic iodine activity The isotopic concurrent iodine spike appearance rates are calculated in Table 7 using the isotopic iodine total activity, Xpurification and Xdecay calculated in this section.

6.3 Post-SGTR Release Rates 6.3.1 Faulted SG - Iodine Release Rates (P-T-S Rupture Flow):

RCS mass released to faulted SG [i.e., SG tube rupture flow] (0 - 0.5 hrs) = 137,246.2 lb RCS mass released to faulted SG [i.e., SG tube rupture flow] (0.5-32 hrs) = 0 lb RCS specific volume @

573 0F saturation temp. = 0.02253 ft3 /Ibm (Ref. 9.14, page 182) 0-0.5 hr P-T-S rupture flow rate from RCS to faulted SG = 137,246.2 lb x 0.02253 ft3/lb = 103.07 cfm 30 minutes 0-0.5 hr iodine P-T-S rupture flow rate from RCS to the faulted SG =

0.5-32 hr iodine P-T-S rupture flow rate from RCS to the faulted SG =

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Mark Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/05/2004 Total mass of steam released from faulted SG to the environment (0 - 0.5 hr) = 56,500 lb (Ref. 9.24, page 8)

Reactor pressure vessel (RPV) average temperature = 588.4 0F (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 SG liquid = 0.02313 ft /lb (Ref. 9.14, page 181)

Iodine partition coefficient = 1.0 (no credit taken for P-T-S rupture flow iodine retention in faulted SG liquid)

Faulted SG steaming rate = 56,500 lb x 0.02313 ft3/lb x (1/1.0) = 43.56 cfm of Faulted SG liquid 30 minutes 0-0.5 hr iodine release from faulted SG to the environment =

if 0.5-32 hr iodine release from faulted SG to the environment = H frriofdGliquid 6.3.2 Intact SGs - Iodine Release Rates (P-T-S Leakage):

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:

P-T-S leakage rate = 1 gallon/minute x (62.4 lb/ft3 x 0.02253 ft3/lb) x 1

ft3 = 0.188 cfm 7.482 gal 0-32 hr iodine P-T-S release rate from RCS to the three intact SGs =

c o

ui Total mass of steam released from intact SGs to the environment (0 - 2 hr) = 465,000 lb (Ref. 9.24, page 8)

Specific volume of SG liquid = 0.02313 ft3/lb Iodine partition coefficient = 1.0 (no credit taken for P-T-S leakage iodine retention in the intact SGs liquid)

Intact SGs steaming rate (0 - 2 hr) = 465,000 lb x 0.02313 ft3/lb x (1/1.0) = 89.63 cfm of Intact SGs liquid 120 minutes Total mass of steam released from intact SG to the environment (2 - 8 hrs) = 1,055,000 lb (Ref. 9.24, page 8)

Intact SGs steaming rate (2 - 8 hr) = 1,055,000 lb x 0.02313 ft3/lb x (1/1.0) = 67.78 cfm of Intact SGs liquid 360 minutes Total mass of steam released from intact SGs to the environment (8 - 24 hrs)

= 1,503,000 lb (Ref. 9.24, page 8)

Intact SGs steaming rate (8 - 24 hr) = 1,503,000 lb x 0.02313 ft3/lb x (1/1.0) = 36.21 cfm of Intact SGs liquid 960 minutes Total mass of steam released from intact SG to the environment (24-32 hrs)

= Total mass of steam released from (24 - 30 hrs) + Total mass of steam released from (30 - 32 lirsg

= 477,000 lb (Ref. 9.24, page 8) + 150,333 lb (Design Input 5.4.5) = 627,333 lb Intact SGs steaming rate (24 - 32 hr) = 627,333 lb x 0.02313 ft3/lb x (1/1.0) = 30.23 cfm of Intact SGs liquid 480 minutes 0 - 2 hr iodine P-T-S leakage release from intact SGs to the environment = PR6 7 q

2 - 8 hr iodine P-T-S leakage release from intact SGs to the environment =

8 - 24 hr iodine P-T-S leakage release from intact SGs to the environment =

t 24 -32 hr iodine P-T-S leakage release from intact SGs to the environment =

Intai INuclear Common Revision 9 Nuclear Common Revision 9 I

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1 l

Mark Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/05/2004 6.3.3 Faulted + Intact SGs - Noble Gas Release Rates:

No noble gas hold-up or decay is modeled in the Faulted and Intact SGs. All RCS mass released to the faulted and intact SGs is treated as if it were released directly to the environment.

Faulted SG - Noble Gas Release Rates:

RCS mass released to environment [i.e., SG tube rupture flow] (0 - 0.5 hrs) = 137,246.2 lb RCS mass released to environment [i.e., SG tube rupture flow] (0.5-32) = 0 lb RCS specific volume @ 573 0F saturation temp. = 0.02253 ft3/lbm (Ref. 9.14, page 182)

P-T-S release rate from RCS to the environment = 137,246.2 lb x 0.02253 ft3/lb = 103.07 cfm 30 minutes 0-0.5 hr noble gas rupture flow rate from RCS to the environ, via the faulted SG =

oIR)5i-0.5-32 hr noble gas rupture flow rate from RCS to the environment via the faulted c

Intact SGs - Noble Gas Release Rates:

RCS P-T-S leakage to the three intact SGs = 1 gallon/minute RCS specific volume ( 573 0F saturation temp. = 0.02253 ft3/lbm (Ref. 9.14, page 182)

P-T-S leakage rate = 1 gallon/minute (62.4 lb/ft3 x 0.02253 ft3/lb) x 1 ft3

= 0.188 cfm 7.482 gal 0-32 hr noble gas P-T-S release rate from RCS to the environ, via the intact SGs =C 0

qu Faulted + Intact SGs - Noble Gas Release Rates:

0-0.5 hr noble gas P-T-S flow rate from RCS to the environ, via the faulted and intact SGs =

103.07 cfm + 0.188 cfm = P7M.c IFFu 0.5-32 hr noble gas P-T-S flow rate from RCS to the environment via the faulted and intact SGs

= 0.0 cfm + 0.188 cfm =rn 6.3.4 Faulted + Intact SGs - Iodine Release Rate - Secondary Liquid The steaming rates for the secondary liquid iodine releases from the fauited and intact steam generators are the sum of those for the faulted and intact SG iodine release rates (Sections 6.3.1 and 6.3.3, respectively), which are then reduced by the iodine partition coefficient of 10 as follows:

0 - 0.5 hr iodine release from all SGs to the environment = (89.63 cfm + 43.56) x (1/10) -

ii 0.5 - 2 hr iodine release from all SGs to the environment = 89.63 cfm x (1/10) =

q quid 2 - 8 hr iodine release from all SGs to the environment = 67.78 cfm x (1/10) = 78 ef liquid 8 - 24 hr iodine release from all SGs to the environment = 36.21 cfm x (1/10) =

62cfm li 24 - 32 hr iodine release from all SGs to the environment = 30.23 cfm x (1/10) =

6.4 Fuel Activity Release Rate The release rate from the source node - fuel to RCS - is calculated such that 99% of the activity in the fuel is released to the RCS in eight hours. The 1% of the activity remaining in the fuel is an insignificant dose I Nuclear Common Revision 9 1 I Nuclear Common Revision 9 I

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Mark Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/05/2004 contribution. The volume of the fuel is inconsequential, given that 99% of the activity will be released irrespective of the fuel volume.

A =AO et Where; AO = Initial Activity in Fuel Node A

Final Activity in Fuel Node X = Removal Rate (vol/hr) t= Removal Time (hr) = 8.0 hr Assuming that 99% of activity is released into the environment, A/AO = 0.01 Therefore, A/Ao =e&t 0.01 = e-8 ln (0.01) = - 82 ln(e)

- 4.605 =-8 X X = - 4.605/-8 = 0.5756 volume/hr Assuming an arbitrary fuel volume of 1,000 cubic feet, the Fuel Release Rate is:

= 0.5756 1/hr x 1,000 ft3 x 1 hr/60 min _ 9.60 ft3/min 6.5 CR 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 1 receives only a fraction of the 2,200 cfm of outside air introduced into the building. The fraction is equivalent to the ratio of the Zone 1 (control room pressure boundary) supply air flow rate [8,000 cfm] to the total control area air conditioning system (CAACS) normal airflow rate [32,600 cfm = 2,200 cfm outside air + 30,400 cfm recirculation air)].

Notes "S" provide the following calculation for the amount of outside air to Zone 1:

= (8,000 cfm / 32,600 cfm) x (2,200 cfm) = 540 cfm Per Notes "S", use 600 cfm for Zone 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 (including 10 percent uncertainty)

CREACS Recirculation Flow Rate With CR Monitors Preferentially Selecting Less Contaminated Air Intake CREACS ventilation flow rate = 8,000 cfm +/- 10% cfm (Ref. 9.6.10)

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. 9.6.9) = 5,000 cfm Nuclear Common Revision 9 l

CALCULATION CONTINUATION SHEET SHEET 25 Of 40 CALC. NO.: S-C-ZZ-MDC-1949

REFERENCE:

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

02/05/2004 0

Mark DruckerfNUCORE, REVIEWER/VERIFIER, DATE 02/05/2004 6.6 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. 9.6.12)

GL 99-02 (Ref. 9.20) requires a safety factor of at least 2 to be used to determine the filter efficiencies to be credited in the design basis accident.

Testing methyl iodide penetration (%) = (100% - i)/safety factor = (100% - il)/2 Where i = charcoal filter efficiency to be credited in the analysis CREACS Charcoal Filter 2.5%= (100%- i)/2 5%= (100%-X1) l = 100% - 5% = 95%

HEPA Filter HEPA filter efficiency = 99% (Ref. 9.6.11). HEPA filter efficiency of 95% is used in the analysis Safety Grade Filter Efficiency Credited (%)

Filter Aerosol Elemental Organic CREACS 95 95 95 6.7 CR Intake Monitor Setpoint Evaluation:

CR Detector Specific Efficiencies:

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

Kr-85 = 2.5 IE8 cpm/gCi/cc (Ref. 9.12, Table 13)

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

Concentration = (RCS activity)/(RCS volume) x (RCS leak rate) x (X/Q)

RCS Activity Xe-133 = 9.16E+04 Ci and Kr-85 = 2.889E+03 Ci (Tables 6 & 9)

RCS Volume = 12,446ft(Ref,-9.6.3) 0-2 hour X/Q = 1.57E:Q2 sec/m3 (Ref. 9.5, Section 8.1)

RCS Leak Rate of Noble Gas isotopes directly to the environment during first 30 minutes

= 103.07 ft3/min (Section 6.3.3) (Assuming the faulted SG released through MSSV Set 1)

Xe-133 Concentration at the CR intake

= (9.16E+04 Ci / 12,446 ft3) x (103.07 ft3/min) x (1.57E-02 sec/M3) x (1/60 min/sec)

= 1.99E-01 Ci/m 3 = 1.92E-01 giCi/cc Kr-85 Concentration at the CR intake I Nuclear Common Revision 9 1 I Nu l a o m nR v s o

CALCULATION CONTINUATION SHEET SHEET 26 of 40 CALC. NO.: S-C-ZZ-MDC-1949

REFERENCE:

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

02/05/2004 0

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

= (2.889E+03 Ci / 12,446 ft3) x (103.07 ft3/min) x (1.57E-02 sec/m3) x (1/60 min/sec)

= 6.27E-03 Ci/m3 = 6.06E-03 jtCi/cc Count Rate Xe-133 = 1.92E-01 jICi/cc x 4.1 1E7 cpm/ptCi/cc = 7.891E+06 cpm Count Rate Kr-85 = 6.06E-03 [iCi/cc x 2.51E8 cpm/nlCi/cc = 1.521E+06 cpm Combined Count Rate of CR Intake Detector = 7.891E+06 cpm + 1.521E+06 cpm

= 9. 412E+06 cpm >> CR Monitor Setpoint of 2,480 cpm The Post-SGTR accident concentration at the CR intake greatly exceeds the CR monitor setpoint value of 2,480 cpm, therefore the CR monitor will respond instantaneously.

I Nuclear Common Revision 9 l Nuclear Common Revision 9

I CALCULATION CONTINUATION SHEET ISHEET 27 of 40 CALC.NO.: S-C-ZZ-MDC-1949

REFERENCE:

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

02/05/2004 0

Mark DruckerfNUCORE, REVIEWER/VERIFIER, DATE 02/05/2004 7.0 RESULTS

SUMMARY

7.1 The results of the SGTR accident with the preaccident iodine spike are summarized in the following table with the CR monitors preferentially selecting the less contaminated CR air intake:

SGTR Accident - Preaccident Iodine Case TEDE Dose (rem)

Receptor Location Control Room EAB LPZ P-T-S Iodine Release 5.05E-01 2.17E+00 3.25E-01 SGTRPIDOO (occurs at t = 0)

SC Liquid Iodine Release 3.83E-03 4.97E-03 1.33E-03 SGTRSLIDOO (occurs at t = 0)

Noble Gas Release 9.28E-03 3.37E-02 4.88E-03 SGTRPNGOO (occurs at t = 0)

Total 5.18E-01 2.21E+00 3.3 lE-01 Allowable TEDE Limit 5.FOE+00 2.50E+01 2.50E+01 7.2 The results of the SGTR accident with the concurrent iodine spike are summarized in the following table with the CR monitors preferentially selecting the less contaminated CR air intake:

SGTR Accident - Concurrent Iodine Case TEDE Dose (rem)

Receptor Location Control Room EAB LPZ Iodine Release 5.72E-01 1.53E+00 3.23E-01 SGTRCIDO1 (occurs at t = 0)

SC Liquid Iodine Release 3.83E-03 4.97E-03 1.33E-03 SGTRSLIDOO (occurs at t = 0)

Noble Gas Release 9.28E-03 3.37E-02 4.88E-03 SGTRPNGOO (occurs at t = 0)

Total 5.85E-01 1.57E+00 3.29E-01 Allowable TEDE Limit 5.OOE+00 2.50E+00 2.50E+00 I Nuclear Common Revision 9 1 Nuclear Common Revision 9

CALCULATION CONTINUATION SHEET SHEET 28 of 40 CALC. NO.: S-C-ZZ-MDC-1949

REFERENCE:

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

02/05/2004 0

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

8.0 CONCLUSION

S:

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

9.0 REFERENCES

1.

U.S. NRC Regulatory Guide 1.1 83, 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-O-ZZ-MCS-0225, Rev.0, RADTRAD Computer Code, Version 3.02

5.

Calculation No. S-C-ZZ-MDC-1959, Rev. 0, CR X/Qs Using ARCON96 Code - Non-LOCA Releases

6.

Salem 1 & 2 Technical Specifications:

1.

Specification 3.4.8, Salem Unit I Limiting Condition for Operation (LCO) for Reactor Coolant System Specific Activity

2.

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

3.

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

4.

Specification 3.4.6.2, Salem Unit 1 LCO for Reactor Coolant System Operational Leakage

5.

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

6.

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

7.

Technical Specification Figure 3.4-1, Salem Unit 1/Unit 2 Dose Equivalent I-131 Primary Coolant Specific Activity Limit

8.

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

9.

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

10.

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

11.

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

12.

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

13.

Specification 1.10, Salem Unit 1/Unit 2 Dose Equivalent I-131 Nuclear Common Revision 9 INuclear Common Revision 9

I CALCULATION CONTINUATION SHEET ISHEET 29 of 40 CALC. NO.: S-C-ZZ-MDC-1949

REFERENCE:

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

02/05/2004 0

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

14.

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

15.

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

16.

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

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. 1 EC-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-93-057, Rev. 0, Salem Coolant Activity, Iodine Spike Appearance Rates and Secondary Activity

14.

ASME Steam Tables, Sixth Edition

15.

Notification No. 20134987, Order 70030162, Operation 0060, Engineering Evaluation Maximum Letdown Flow

16.

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

17.

Westinghouse Calculation No. CN-CRA-93-087, Rev. 0, Salem Steam Generator Tube Rupture Offsite Dose Analysis

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.

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

21.

Not Used.

22.

Not Used.

23.

GE-NE-0000-001 1-3853-R1; DRF 0000-0004-6923, Revision 1, Class III, June 2003, Project Task Report T0807, Coolant Radiation Sources

24.

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

25.

Letter from Huckabee, J (Westinghouse), to Rosenfeld, E.S. (PSE&G) "Evaluation of SGTR analysis using assumed 30 minute operator action time," Letter No. PSE-96-579, March 7, 1996, [NFSI-96-102 (Roll 09134 Frame 1531)]

INuclear Common Revision 9 Nuclear Common Revision 9

CALCULATION CONTINUATION SHEET SHEET 30 of 40 CALC. NO.: S-C-ZZ-MDC-1949

REFERENCE:

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

02/05/2004 0

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

Table I Iodine Isotopic Dose Conversion Factors Isotopic Conversion Iodine Dose Factor Dose Isotope Conversion Conversion Factor Factor (Sv/Bq)

(rcm/CiISv/Bq)

(rem/Ci)

A B

C=AxB 1-131 2.92E-07 3.70E+12 1.08E+06 1-132 1.74E-09 3.70E+12 6.44E+03 I-133 4.86E-08 3.70E+12 1.80E+05 1-134 2.88E-10 3.70E+12 1.07E+03 1-135 8.46E-09 3.70E+12 3.13E+04 A From Reference 9.7, Page 136 Table 2 Iodine Scaling Factors Preaccident Iodine Spike & Equilibrium Iodine Concentration 1% Failed Fuel Iodine Iodine Dose Product Isotope Activity Conversion Concentration Factor piCi.rem/Ci.g (4Ci/g)

(rem/Ci)

(rem)

A B

(A x B) 1-131 2.80E+00 1.08E+06 3.02E+06 1-132 2.80E+00 6.44E+03 1.80E+04 1-133 4.20E+00 1.80E+05 7.55E+05 1-134 5.70E-01 1.07E+03 6.08E+02 1-135 2.301E+00 3.133E+04 7.20E+04 Total

--3.87E+06 A From Reference 9.11, Table 4; and B from Table 1. _.

1-131 DE Based on 1% FF Iodine Concentration I 3.58E+00 Ilodine Scaling Factor Based on 1.0 ptCi/g DE 1-1311 2.791E-01 l Nuclear Common Revision 9 l

I ~

I CALCULATION CONTINUATION SHEET ISHEET 31 of 40 CALC. NO.: S-C-ZZ-MDC-1949

REFERENCE:

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

02105/2004 0

Mark Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/05f2004 Table 3 RCS Iodine Concentration Based on 1.0 pCi/g DE 1-131 1% Failed Fuel Iodine 1.0 pCi/g Iodine Scaling DE 1-131 Isotope Activity Factor Activity Concentration 1.0.Ci/g Concentration (PCilg)

DE 1-131 (pCi/g)

A B

C=AxB 1-131 2.80E+00 2.791E-01 7.814E-01 I-132 2.80E+00 2.791E-01 7.814E-01 1-133 4.20E+00 2.791E-01 1.172E+00 1-134 5.70E-01 2.791E-01 1.5911E-01 1-135 2.30E+00 2.791E-01 6.419E-01 A From Reference 9.11, Table 4 B Scaling Factor Based on 1.0 piCi/g DE 1-131 From Table 2 Table 4 Total RCS Iodine Activity - Preaccident Iodine Spike Case Tech Spec Reactor Total RCS Activity RCS Coolant 1.0 PCig 60.0 pCi/g Isotope Activity Mass DE 1-131 DE 1-131 Concentration (ItCi/g)

(g)

(Ci)

(Ci)

A B

C=AxB/IE-6 Cx6O 1-131 7.814E-01 3.523E+08 2.753E+02 1.652E+04 1-132 7.814E-01 3.523E+08 2.753E+02 1.652E+04 1-133 1.172E+00 3.523E+08 4.130E+02 2.478E+04 1-134 1.591 E-01 3.523E+08 5.604E+0 1 3.363E+03 1-135 6.419E-01 3.523E+08 2.261E+02 1.357E+04 A From Table 3 B Reactor coolant mass from Section 6.1 I Nuclear Common Revision 9 1 Nuclear Common Revision 9

I I F

CALCULATION CONTINUATION SHEET ISHEET 32 of 40 CALC. NO.: S-C-ZZ-MDC-1 949

REFERENCE:

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

02/05/2004 0

Mark Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/05/2004 Table 5 Total RCS Noble Gas Activity - Preaccident Iodine Spike 1% Failed Fuel Noble Gas Primary Total Isotope Activity Coolant Noble Gas Concentration Mass Activity (PCi/g)

(g)

(Ci)

A B

C=AxB/1E6 Kr-83m 4.OOOE-01 3.523E+08 1.409E+02 Kr-85m 1.700E+00 3.523E+08 5.989E+02 Kr-85 8.200E+00 3.523E+08 2.889E+03 Kr-87 1.OOOE+00 3.523E+08 3.523E+02 Kr-88 3.000E+00 3.523E+08 L.057E+03 Xe-131 m 2.100E+00 3.523E+08 7.398E+02 Xe-133m 1.700E+01 3.523E+08 5.989E+03 Xe-133 2.600E+02 3.523E+08 9.160E+04 Xe-135m 4.900E-01 3.523E+08 1.726E+02 Xe-135 8.500E+00 3.523E+08 2.995E+03 Xe-138 6.100E-01 3.523E+08 2.149E+02 A From Reference 9.11, Table 4 B From Section 6.1 I Nuclear Common Revision 9 1 Nuclear Common Revision 9

I CALCULATION CONTINUATION SHEET ISHEET 33 of 40 CALC. NO.: S-C-ZZ-MDC-1949

REFERENCE:

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

02/05/2004 0

Mark Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/05/2004 Total Table 6 Iodine & Noble Gas Activities in RCS Preaccident Iodine Snike Iodine &

RADTRAD Noble Gas Nuclide Isotope Total Inventory Activity File (Ci)

(Ci) 1-131 1.652E+04

.1652E+05 1-132 1.652E+04

.1652E+05 1-133 2.478E+04

.2478E+05 1-134 3.363E+03

.3363E+04 1-135 1.357E+04

.1357E+05 Kr-83m 1.409E+02

.1409E+03 Kr-85m 5.989E+02

.5989E+03 Kr-85 2.889E+03

.2889E+04 Kr-87 3.523E+02

.3523E+03 Kr-88 1.057E+03

.1057E+04 Xe-131m 7.398E+02

.7398E+03 Xe-133m 5.989E+03

.5989E+04 Xe-133 9.160E+04

.9160E+05 Xe-135m 1.726E+02

.1726E+03 Xe-135 2.995E+03

.2995E+04 Xe-138 2.149E+02

.2149E+03 Iodine Activity from Table 4 Noble gas activity from Table 5 Nuclear Common Revision 9 I Nuclear Common Revision 9 1

I CALCULATION CONTINUATION SHEET ISHEET 34 of 40 CALC.NO.: S-C-ZZ-MDC-1949

REFERENCE:

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

02/05/2004 0

Mark Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/05/2004 Table 7 Iodine Appearance Rate for Concurrent Iodine Spike 1.0 AiCi/g Letdown Iodine Iodine DE 1-131 Decay Purification Appearance Spike Isotope Iodine Constant Removal Rate Appearance Activity A

Rate Rate (Ci)

(sec) '

(sec)-'

(Ci/sec)

(Ci/sec)

A B

C D=Ax(B+C)

E=Dx335 1-131 2.753E+02 9.977E-07 3.686E-05 1.042E-02 3.491 1-132 2.753E+02 8.426E-05 3.686E-05 3.334E-02 11.170 1-133 4.130E+02 9.257E-06 3.686E-05 1.904E-02 6.380 I-134 5.604E+01 2.196E-04 3.686E-05 1.437E-02 4.815 1-135 2.2611E+02 2.924E-05 3.686E-05 1.495E-02 5.008 Total Appearance Rate (Ci/sec) 30.864 A From Table 4 B From Reference 9.23, Appendix A (A per hr is converted into per sec)

C From Section 6.2 Table 8 Total Iodine Activity In Fuel - Concurrent Iodine Spike Iodine Concurrent Concurrent Spike Iodine Spike Iodine Spike Isotope Appearance Duration Total Rate Iodine Activity (Ci/sec)

(hrs)

(Ci)

A B

AxBx3600 I-131 3.491 8

1.006E+05 1-132 11.170 8

3.217E+05 I-133 6.380 8

1.837E+05 I-134 4.815 8

1.387E+05 I-135 5.008 8

1.442E+05 A From Table 7 B From Reference 9.1, Appendix F, Section 2.2 I Nuclear Common Revision 9 l Nuclear Common Revision 9

I CALCULATION CONTINUATION SHEET ISHEET 35 of 40 CALC. NO.: S-C-ZZ-MDC-1949

REFERENCE:

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

02/05/2004 0

Mark Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/05/2004 Table 9 Total Iodine & Noble Gas Activities In Fuel Concurrent Iodine Spike Iodine &

RADTRAD Noble Gas Nuclide Isotope Total Inventory Activity File (Ci)

(Ci) 1-131 1.006E+05

.1006E+06 1-132 3.217E+05

.3217E+06 1-133 1.837E+05

.1837E+06 1-134 1.387E+05

.1387E+06 1-135 1.442E+05

.1442E+06 Kr-83m 1.409E+02

.1409E+03 Kr-85m 5.989E+02

.5989E+03 Kr-85 2.889E+03

.2889E+04 Kr-87 3.523E+02

.3523E+03 Kr-88 1.057E+03

.1057E+04 Xe-131m 7.398E+02

.7398E+03 Xe-133m 5.989E+03

.5989E+04 Xe-133 9.160E+04

.9160E+05 Xe-135m 1.726E+02

.1726E+03 Xe-135 2.995E+03

.2995E+04 Xe-138 2.149E+02

.2149E+03 A - Iodine activity from Table 8 A - Noble gas from Table 6 INuclear Common Revision 9 Nuclear Common Revision 9

I CALCULATION CONTINUATION SHEET ISHEET 36 of 40 CALC.NO.: S-C-ZZ-MDC-1949

REFERENCE:

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

02/05/2004 0

Mark Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/05/2004 Table 10 Secondary Side Iodine Activity 1.0 pCi/g Tech Spec Coolant Mass Secondary RADTRAD DE 1-131 Secondary In All Four Side Nuclide Isotope Activity Coolant Steam Activity Inventory Concentration Activity Limit Generators File (11Clg)

(OCig)

(g)

(Ci)

(Ci)

A B

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

.1810E+02 1-132 7.814E-01 0.10 2.316E+08 18.098

.1810E+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 3 B From Reference 9.6.6 C From Section 6.1 I Nuclear Common Revision 9 l Nuclear Common Revision 9

T CALCULATION CONTINUATION SHEET l SHEET 37 of 40 CALC. NO.: S-C-ZZ-MDC-1949

REFERENCE:

LCR S03-0S Gopal PateIINUCORE, ORIGINATOR, DATE REV:

02/05/2004 0

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

Figure 1: RADTRAD Nodalization For SGTR Accident with a Preaccident Iodine Spike Release P-T-S Leakage Iodine Release Path Noble Gas Release Path

........ a.

Secondary Liquid Iodine Release Rates per Section 6.3.4 Total SG volume of 8,182 ft3 is used for secondary liquid iodine release INuclear Common Revision 9 l Nuclear Common Revision 9

I I

CALCULATION CONTINUATION SHEET ISHEET 38 of 40 CALC. NO.: S-C-ZZ-MDC-1949

REFERENCE:

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

02/05/2004 0

l f

Mark Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/05/2004 Figure 2: RADTRAD Nodalization For SGTR Accident with a Concurrent Iodine Spike Release P-T-S Leakage Iodine Release Path E~-Noble Gas Release Path Secondary Liquid Iodine Release Rates per Section 6.3.4 Total SG volume of 8,182 ft3 is used for secondary liquid iodine release Nuclear Common Revision 9

I CALCULATION CONTINUATION SHEET I SHEET 39 of 40 CALC. NO.: S-C-ZZ-MDC-1949

REFERENCE:

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

02/05/2004 0

l l

Mark Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/05/2004 Figure 3 - RADTRAD Nodalization For CR Rcsponsc With CR HVAC Intake Radiation Monitors Preferentially Selecting the Less Contaminated CR Air Intake i

I Nuclear Common Revision 9 l Nuclear Common Revision 9

I CALCULATION CONTINUATION SHEET SHEET 40 of 40 CALC. NO.: S-C-ZZ-MDC-1949

REFERENCE:

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

02/05/2004 0

Mark Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/05/2004 12.0 AFFECTED DOCUMENTS:

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

Document to be voided:

Vendor Technical Document No. 322260, Rev. 4, Radiological Dose Consequence At the EAB/LPZ And In The Control Room With Modified Control Room Ventilation Design - Steam Generator Tube Rupture.

Documents to be revised:

UFSAR Section 15.4.4, Steam Generator Tube Rupture UFSAR Table 1 5.4-7B, Steam Generator Tube Rupture Parameters and Assumptions UFSAR Table 15.4-7C, Steam Generator Tube Rupture Dose Consequences UFSAR Table 15.4-8, Primary and Secondary Coolant Iodine and Noble Gas Activities UFSAR Table 15.4-9, Breathing Rates 13.0 ATTACHMENTS:

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

Calculation No: S-C-ZZ-MDC-1 949, Rev. OIR2 Comment Resolution Form 2 - Mark Drucker Certification for Design Verification Form-I RCPD Form-I I Nuclear Common Revision 9 l Nuclear Common Revision 9

CALCULATION CONTINUATION SHEET

[SHEET 40 of 40 CALC. NO.: S-C-ZZ-MDC-1949

REFERENCE:

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

02/05/2004 0

Mark Drucker/NUCORE, REVIEWER/VERIFIER, DATE 02/05/2004 12.0 AFFECTED DOCUMENTS:

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

Document to be voided:

Vendor Technical Document No. 322260, Rev. 4, Radiological Dose Consequence At the EAB/LPZ And In The Control Room With Modified Control Room Ventilation Design - Steam Generator Tube Rupture.

Documents to be revised:

UFSAR Section 15.4.4, Steam Generator Tube Rupture UFSAR Table 15.4-7B, Steam Generator Tube Rupture Parameters and Assumptions UFSAR Table 15.4-7C, Steam Generator Tube Rupture Dose Consequences UFSAR Table 15.4-8, Primary and Secondary Coolant Iodine and Noble Gas Activities UFSAR Table 15.4-9, Breathing Rates 13.0 ATTACHMENTS:

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

Calculation No: S-C-ZZ-MDC-1 949, Rev. OIR2 Comment Resolution Form 2 - Mark Drucker Certification for Design Verification Form-1 RCPD Form-i l Nuclear Common Revision 9 I Nuclear Common Revision 9 I

b Attachment A S-C-ZZ-MDC-1949, Rev. 0 2 Diskettes With Various Electronic Files