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Calculation - DC00040-119 - Reactor Coolant Pump Locked Rotor - TSC
	ML14324A323
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Site:	Summer
Issue date:	07/30/2012
From:	Feehan M; South Carolina Electric & Gas Co
To:	; Office of Nuclear Reactor Regulation
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ML14324A264	List: ML14324A217; ML14324A323; ML14324A324; ML14324A325; ML14324A327; RC-14-0179, Calculation - DC00040-119 - Reactor Coolant Pump Locked Rotor - TSC; RC-14-0179, Calculation - DC00040-120 - Fuel Handling Accident -TSC; RC-14-0179, Calculation - DC00040-121 - Steam Generator Tube Rupture - TSC; RC-14-0179, Calculation - DC00040-122 - Loss of Coolant Accident - TSC; RC-14-0179, Calculation - DC00040-123 - Rod Ejection Accident - TSC; RC-14-0179, Virgil C. Summer Nuclear Station (VCSNS) Unit 1 - License Amendment Request LAR-13-02396 Request for NRC Approval of Changes to the Radiation Emergency Plan;
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	LAR-13-02396, RC-14-0179 DC00040-119, Rev. 0
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ES-0412 ATTACHMENT I PAGE 1 OF 2 REVISION 5 Subject Code SOUTH CAROLINA ELECTRIC AND GAS COMPANY 004 CALCULATION RECORD Page 1 of 70 Calculation Title Calculation Number Revision Status Reactor Coolant Pump Locked Rotor -

D000040-119 0

A TSC Parent Document System Safety Class

[]Partial Calc. Revision ECR 50786 N/A q NN q QR SR Complete Calc. Revision Originator Discipline Organization Date XREF Number Michael Feehan AE WorleyParsons 07/30/2012 N/A CALCULATION INFORMATION Content

Description:

The calculation determines the impact of the Alternate Source Term (AST) on the new Nuclear Operations Building (NOB) VC Summer Unit 1 Technical Support Center (TSC) dose following a postulated Reactor Coolant Pump Locked Rotor Accident. This analysis is performed in accordance with the requirements of Appendix G of the USNRC Regulatory Guide 1.183.

Affected Components/Calculations/Documents: FSAR Section 15.4.4.

Piping Reconciliation Completed per QA-CAR-0089-18: []This Revision q Previous Revision N/A Contains Preliminary Data/Assumptions:

No q Yes, Affected Pages:

Computer Program Used:

q No Yes, Validated per WorleyParsons computer program validation process (others) vendors name q Yes, Validated in accordance with SAP-1040/ES-413 (ref. 3.4 & 3.5) q Yes, Validated [ES-0412]

q Computer Program Validation Calculation VERIFICATION q Continued, Attachment Scope:

Verify input, methodology, and output and assure that the calculation is in compliance with ES-412 and addresses the work scope and deliverables documented in CGGS-1 1-114.

Verifier: Michael M. Waselus Assigned by: Paul L. Bunker&4,7150112-Engineering Personnel /Date Owne 's cceptance Review Verifier/Date Responsible Engi eer/Date Required for all engineering work performed by contractor personnel not enrolled in the VCSNS Engineering Training Program RECORDS l

l ^ l2 To Records Mgmt: At kN Approval/Date Initials/Date A

Distribution: Calc File (Original)

ES-0412 ATTACHMENT I PAGE 2 OF 2 REVISION 5 SOUTH CAROLINA ELECTRIC & GAS COMPANY REVISION

SUMMARY

Calculation Number DC00040-119 Revision Number.

Summary Description Page 2 of 70 0

Initial Revision

DO00040-119, Revision 0 PAGE 3 OF 70 TABLE OF CONTENTS DESCRIPTION PAGE

1.0 INTRODUCTION

.................................................................................................................. 5 1.1 Purpose......................................................

......... 5 1.2 Objective................................................................................................................................. 5 1.3 Acceptance Criteria................................................................................................................. 5 1.4 Approach................................................................................................................................. 6 2.0 ASSUMPTIONS..................................................................................................................... 7 3.0 COMPUTER CODES............................................................................................................. 8

4.0 REFERENCES

....................................................................................................................... 9 5.0 DESIGN INPUTS.................................................................................................................11 6.0 METHODOLOGY...............................................................................................................16 7.0 COMPUTATIONS............................................................................................................... 17 7.1 Conservative Case................................................................................................................. 17 7.1.1 Calculation of Activity Released to Environment................................................................ 17 7.1.2 RADTRAD Model - Input Description................................................................................ 21 8.0

SUMMARY

OF RESULTS................................................................................................. 23 9.0 DISPOSITION OF RESULTS............................................................................................. 23

D000040-119, Revision 0 PAGE 4 OF 70 TABLE OF CONTENTS ATT. #

DESCRIPTION PAGE 1

RADTRAD Model.................................................................................................................24 2

Conservative Source Terms...................................................................................................26 3

RADTRAD Source Input File for Conservative Case...........................................................37 4

RADTRAD Release Fraction Files........................................................................................48 5

Conservative Case RADTRAD Input/Output Files...............................................................50 6

Westinghouse Transmittal #CGE-93-0007SGUL (PS-CGE-0807) dated 3/1/93..................68 TABLE #

DESCRIPTION PAGE 1

Core Activity atT=0............................................................................................................11 2

Reactor Coolant Fission and Corrosion Specific Activity at Equilibrium.............................12 3

Fraction of Fission Product Inventory in the Gap Available for Release..............................12 4

TSC & Offsite Atmospheric Dispersion Factors...................................................................14 5

TSC & Offsite Breathing Rates.............................................................................................15 6

TSC Occupancy Factors........................................................................................................15 7

Isotopic Decay Constants.......................................................................................................15 8

RCPLR Dose in Rems TEDE - TSC......................................................................................23

D000040-119, Revision 0 PAGE 5 OF 70

1.0 INTRODUCTION

1.1 PURPOSE The purpose of this calculation is to assess the impact of the Alternative Source Term (AST) on the Reactor Coolant Pump Locked Rotor accident. This analysis was performed in accordance with the requirements of USNRC Regulatory Guide 1.183 (Reference 4.3). As a result of the accident, 15% of the fuel rods in the core are considered to fail and their gap activity is released to, and simultaneously mixed with, the reactor coolant.

A conservative analysis of the potential Technical Support Center (TSC) dose resulting from a reactor coolant pump locked rotor is analyzed assuming all loops are operating, with one loop experiencing a locked rotor accident.

At the beginning of the postulated locked rotor accident, the plant is assumed to be in operation under the most adverse steady-state operating conditions, i.e., maximum steady-state power level, maximum steady-state pressure, and maximum steady-state coolant average temperature. This analysis incorporates assumptions of 1 percent defective fuel and steam generator leakage prior to the postulated accident for a time sufficient to establish equilibrium specific activity levels in the secondary system.

No credit is taken for the pressure reducing effect of the pressurizer relief valves, pressurizer spray, steam dump or controlled feedwater flow after plant trip. Although these operations are expected to occur and would result in a lower peak pressure, an additional degree of conservatism is provided by ignoring their effect.

1.2 OBJECTIVE The objective of this analysis is to demonstrate that the VCSNS design features, following the Reactor Coolant Pump Locked Rotor accident, provide sufficient margin to ensure that the post accident TSC dose satisfy the I OCFR50.67 Accident Source Term criteria (Reference 4.1), 10CFR50 Appendix A, General Design Criteria (GDC) 19 (Reference 4.2), and the guidance in USNRC Regulatory Guide (RG) 1.183 (Reference 4.3).

1.3 ACCEPTANCE CRITERIA The TSC is to provide direct management and technical support to the control room during an accident; consequently it shall have the same radiological habitability as the control room under accident conditions. TSC personnel shall be protected from radiological hazards, including direct radiation and airborne radioactivity from in-plant sources under accident conditions, to the same degree as control room personnel.

The TSC acceptance criteria for the radiological consequences of this accident are the same as the control room radiation exposures; within the l OCFR50.67 limits; specifically:

Adequate radiation protection is provided to permit access to and occupancy of the TSC under accident conditions without personnel receiving radiation exposures in excess of 0.05 Sv (5 rem) total effective dose equivalent (TEDE) for the duration of the accident.

D000040-119, Revision 0 PAGE 6 OF 70 1.4 APPROACH To demonstrate the margin provided by the plant's design features in complying with the regulatory guidelines and requirements, a design basis analysis for the Reactor Coolant Pump Locked Rotor accident is evaluated. The dose models used to calculate the offsite and TSC radiological consequences are based on those presented in RG 1.183. The RADTRAD computer code (Reference 4.4) is used to calculate the offsite and TSC doses. This code was developed for the NRC and is used to calculate radiation doses by simulating the movement of radioactivity through various regions of the Containment, its removal by various processes, and its leakage to the environment. The protected TSC operator doses from the inleakage of radioactive materials into the control building are calculated based upon a whole body geometry factor as described by Murphy-Campe (Reference 4.5).

D000040-119, Revision 0 PAGE 7 OF 70 2.0 ASSUMPTIONS 2.1 The analysis is consistent with the guidance provided in USNRC Regulatory Guide 1.183 (Reference 4.3), Appendix G.

2.2 Acceptable assumptions regarding core inventory and the release of radionuclides from the fuel are provided in Regulatory Position 3 and Appendix G of Reference 4.3. The fraction of fission product inventory in the gap available for release due to fuel breach is given in Table 3 of Reference 4.3. The activity released from the fuel should be assumed to be released instantaneously and homogeneously through the primary coolant (Reference 4.3, Appendix G.3).

2.3 The chemical form of the radioiodine released from the damaged fuel should be assumed to be 95% cesium iodide (CsI), 4.85% elemental iodine, and 0.15% organic iodide. Iodine releases from the steam generators to the environment should be assumed to be 97%

elemental and 3 % organic. These fractions apply to iodine released as a result of fuel damage and to equilibrium iodine concentrations in the Reactor Coolant and Secondary Coolant Systems (Reference 4.3, Appendix G.4).

2.4 The primary-to-secondary leak rate in the steam generators should be assumed to be the leak-rate-limiting conditions for operation specified in the technical specifications. The leakage should be apportioned between the steam generators in such a manner that the calculated dose is maximized (Reference 4.3, Appendix G.5.1). For this event, the locked rotor does not affect any one loop/SG differently than the other. Consequently, it is assumed that each loop has an equal leak rate.

2.5 The density used in converting volumetric leak rates (e.g., gpm) to mass leak rates (e.g., lb/hr) should be consistent with the basis of surveillance tests used to show compliance with leak rate technical specifications. These tests are typically based on cool liquid (Reference 4.3, Appendix G.5.2). Facility instrumentation used to determine leakage is typically located on lines containing cool liquids. In most cases, the sensitivity should be assumed to be 1.0 gm/cc (62.4 lb/ft3).

2.6 The primary-to-secondary leakage should be 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 1000 C (212° F). The release of radioactivity should be assumed to continue until shutdown cooling is in operation and releases from the steam generators have been terminated (Reference 4.3, Appendix G.5.3).

2.7 The release of fission products from the secondary system should be evaluated with the assumption of a coincident loss of offsite power (Reference 4.3, Appendix G.5.4).

2.8 All noble gas radionuclides released from the primary system are assumed to be released to the environment without reduction or mitigation (Reference 4.3, Appendix G.5.5).

2.9 The iodine and particulate transport model for release from the steam generators is as follows (Reference 4.3, Appendix G.5.6, and Appendices E.5 & E.6). Please note, since there is no steam generator failure, portions of Appendix E.5 are not applicable. Section E.6 does not apply because steam generator tube uncovery does not occur.

DC00040-119, Revision 0 PAGE 8 OF 70 The primary-to-secondary leakage to the SGs is assumed to mix instantaneously and homogeneously with the secondary water without flashing.

The radioactivity in the secondary water is assumed to become a vapor at a rate that is the function of the steaming rate and the partition coefficient. A partition coefficient for iodine of 100 may be assumed. Although Reference 4.3, Appendix E, footnote 3 defines the partition coefficient in terms of 12, or elemental iodine, it is assumed that the factor of 100 is an overall partition coefficient, applicable to all iodine species. It is conservatively assumed that the partition coefficient for the alkali metals is 100, the same as the halogens.

2.10 Consistent with Section 3.6 of Reference 4.3, the amount of fuel damage caused by non-LOCA events is analyzed to determine, for the case resulting in the highest radioactivity release, the fraction of the fuel that reaches or exceeds the initiation temperature of fuel melt and the fraction of fuel elements for which the fuel clad is breached. For the Reactor Coolant Pump Locked Rotor event, Table 3 of RG 1.183 (Reference 4.3) specifies noble gas, alkali metal, and iodine fuel gap release fractions for the breached fuel.

2.11 No credit is taken for the pressure reducing effect of the pressurizer relief valves, pressurizer spray, steam dump or controlled feedwater flow after plant trip 2.12 It is assumed that at the time of reactor trip, there is a loss of offsite power. This drives the release from the secondary coolant system through the steam generator atmospheric relief valves or safety valves, since condenser cooling is lost.

2.13 Isotopes considered in the AST radiological consequence analyses are restricted to the 60 isotopes addressed in the RADTRAD Computer Code (Reference 4.4).

2.14 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after the accident, the Residual Heat Removal System starts operation to cool down the plant. At this point, no steam and activity are released to the environment.

2.15 Per Reference 4.3, Table 3, the non-LOCA fraction of fission products inventory in the gap is acceptable for use if the peak fuel burnup does not exceed 62,000 MWD/MTU and the maximum linear heat generation rate does not exceed 6.3 kw/ft. peak rod average power for burnups exceeding 54 GWD/MTU. To account for possible variation in burnup and rod power, the Table 3 non-LOCA fraction of fission products inventory in the gap is conservatively doubled based on Reference 4.16.

2.16 It is assumed that the TSC does not isolate for the duration of the accident. An assumed inflow of 13,000 cfm conservatively bounds the expected design flow rate of 12,000 cfm.

The doses are conservatively calculated assuming no credit for filtration for the duration of the accident.

2.17 It is assumed that the mass and associated activity released for various time periods is linear.

3.0 COMPUTER CODES 3.1 RADTRAD 3.03 verified by WorleyParsons 5/27/2003.

D000040-119, Revision 0 PAGE 9 OF 70

4.0 REFERENCES

4.1 Title 10, Code of Federal Regulations Part 50.67, "Accident Source Term."

4.2 Title 10 Code of Federal Regulations Part 50 Appendix A, "General Design Criteria for Nuclear Power Plants", Criterion 19.

4.3 USNRC Regulatory Guide 1.183, July 2000, "Alternate Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors."

4.4 NUREG/TSC-6604, "RADTRAD: A Simplified Model for Radionuclide Transport and Removal and Dose Estimation", December 1997 and NUREG/TSC-6604 (SAND98-0272/1),

Supplements 1 & 2, "RADTRAD: A Simplified Model for Radionuclide Transport and Removal and Dose Estimation", June 8, 1999 & October 2002.

4.5 "Nuclear Power Plant Control Room Ventilation System Design for Meeting General Design Criteria GDC 19", K. G. Murphy and K. M. Campe, USAEC, 13th AEC Air Cleaning Conference, August 1974.

4.6 VCSNS Calculation D000040-111, Revision 0, "Short Term Accident X/Qs."

4.7 VCSNS Calculation D000040-079, Revision 3, "Atmospheric Dispersion Coefficients for Control Room."

4.8 Westinghouse Transmittal #CGE-93-0007SGUL (PS-CGE-0807) dated 3/1/93, Transmittal of VCSNS RSG/Uprating: Steam Release for Dose Analysis and Write-up for Source Terms (Attachment 6).

4.9 USNRC Regulatory Guide 1.49, "Power Levels of Nuclear Power Plants", Revision 1.

4.10 GMK Associates, Inc. drawings for VC Summer Nuclear Operations Building.

a.

I MS-82-202 (A2.0) - "Basement Floor Plan", RO b.

I MS-82-213 (A4.0) - "Building Sections", RO c.

1MS-82-700 (M2.0) - "Basement Plan HVAC Ductwork", RO d.

1MS-82-717 (M7.2) - "HVAC Schedules", RO 4.11 VC Summer Nuclear Operations Building, A/E Project #07023.05, "Project Manual Including Specifications Issued for Construction", GMK Associates, Inc., May 6, 2011.

4.12 VCSNS Calculation D000040-103, Revision 0, "Calculation of Secondary Side Extended Mass Releases During Cooldown for AST Analyses."

4.13 VCSNS Calculation D000040-005, Revision 4, "Secondary Coolant System Equilibrium Analysis."

4.14 Westinghouse Radiation Analysis Manual, Revision I for VCSNS Uprating updated 10/98 (attachment to letter CGE-98-036).

D000040-119, Revision 0 PAGE 10 OF 70 4.15 "Radioactive Decay Data Tables - A Handbook of Decay Data for Application to Radiation Dosimetry and Radiological Assessments", David C. Kocher, 1981.

4.16 Byron Station, Unit Nos. land 2, and Braidwood Station, Unit Nos. 1 and 2 - Issuance of Amendments RE: Alternate Source Term (TAC nos. MC6221, MC6222, MC6223, and MC6224), issued September 8, 2006.

4.17 VCSNS Design Basis document for Steam Generator Blowdown and Nuclear Blowdown Processing (BD/NB), Revision 3.

4.18 NUREG-1465, "Accident Source Terms for Light-Water Nuclear Power Plants", February 1995.

4.19 VC Summer Calculation D000040-100, "Reactor Coolant Pump Locked Rotor - AST",

Revision 1.

DO00040-119, Revision 0 PAGE 11 OF 70 5.0 DESIGN INPUTS 5.1 Consistent with RG 1.183, Section 3.1 and Appendix G.1 (Reference 4.3) and Reference 4.9, the AST Locked Rotor dose analysis is performed at 102% of the core thermal power level (1.02 x the Original License Thermal Power of 2900 MWth), or 2958 MWth (Reference 4.14).

5.2 The core inventory of the radionuclide groups required for non-LOCA events, based on Tables 3 and 5 of Regulatory Position 3.0 of Reference 4.3, at 102% of the core thermal power is obtained from Reference 4.14 and reproduced below in Table No. 1. Please note, only the noble gas, halogen, and alkali metal isotopes required per References 4.3, 4.4, and 4.18 are included. Core inventories for the design basis source term are taken from Reference 4.14, Table 5-9. Iodine activity is increased by a factor of 2 to account for the Reference 4.14 TID-14844 release fraction.

Table 1: Core Activity at T = O(A)

Isotope Activity (Ci)

Isotope Activity (0)

Kr-85 8.30E+05 Xe-133 1.70E+08 Kr-85m 2.72E+07 Xe-135 3.70E+07 Kr-87 4.96E+07 Cs-134 1.01E+07 Kr-88 6.71E+07 Cs-136 3.08E+06 Rb-86 4.43E+04 Cs-137 5.66E+06 1-131 8.20E+07 1-132 1.20E+08 1-133 1.68E+08 1-134 1.80E+08 I 135 1.54E+08 (A). Please note, for conservative reasons, for all isotopes not included in Reference 4.14, the default core inventories from Table 1.4.3.2-2 of Reference 4.4, corrected to a core thermal power of 2958MWt, are included in this analysis. In addition, the noble gas and iodine core inventories from Reference 4.14 and the corrected core inventories from Reference 4.4 were compared and the larger of the two concentrations is used in this analysis.

D000040-119, Revision 0 PAGE 12 OF 70 5.3 Reactor coolant equilibrium fission and corrosion product specific activity, based on I% fuel defects, is taken from Reference 4.14, Table 5-15 and summarized below for the applicable isotopes.

Table 2: Reactor Coolant Fission and Corrosion Product Specific Activity at Equilibrium Isotope Activity (ItCi/gm)

Isotope Activity Ci/ m Kr-85 7.6E+00 Xe-133 2.9E+02 Kr-85m 1.8E+00 Xe-135 8.6E+00 Kr-87 1.1E+00 Cs-134 4.4E+00 Kr-88 3.2E+00 Cs-136 4.5E+00 Rb-86 3.6E-02 Cs-137 2.IE+00 1-131 3.0E+00 1-132 3.1E+00 1-133 4.6E+00 1-134 6.0E-01 1-135 2.4E+00 5.4 The release fraction from the breached fuel is based on Regulatory Position 3.2, Table 3 (reproduced below as Table No. 3) and Note 11 of Reference 4.3. Per Section 2.15, the release fractions are conservatively doubled. The percent of the fuel rods breached is 15 percent per Reference 4.8. The release fractions from Table 3 are used in conjunction with the fission product inventory calculated with the maximum core radial peaking factor. The maximum radial peaking factor, which is also referred to as the peak integral pin power channel factor or nuclear enthalpy rise hot channel factor, for the core is 1.7, Reference 4.14 Table 3: Fraction of Fission Product Inventory in the Gap Available for Release')

Group Fraction Corrected Fraction 1-131 0.08 0.16 Kr-85 0.10 0.20 Other Noble Gases 0.05 0.10 Other Halogens 0.05 0.10 Alkali Metals 0.12 0.24 See Section 2.15 5.5 Per References 4.8 & 4.12:

Steam released from the 3 steam generators is:

447,900 lbs (0 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ) 868,300 lbs (2 - 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ) 1,200,000 lbs (8 - 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ) 5.6 The dose acceptance criterion (References 4.1 & 4.3) is 5 Rem TEDE for the TSC.

DO00040-119, Revision 0 PAGE 13 OF 70 5.7 Per Reference 4.14, the mass of the reactor coolant system is 1.8E+08 grams or 4.0E+05 lbs (1.8E+08 gm

I lb/454 gm).

5.8 Per Reference 4.14, Table 7.1 the total water mass of the three steam generators is 340,000 lbs.

5.9 Steam generator blowdown flow rates are as follows:

Per Reference 4.17, Section 3.1.1, the minimum blowdown flow rate is 30 gpm per steam generator or 12,756 lb/hr.

5.10 The VCSNS TSC design features include post accident isolation with filtered supply and pressurization. The following parameters are utilized in accessing the post-accident dose consequences to TSC personnel.

The total free volume of the TSC envelope located in the basement of the new Nuclear Operations Building (elevation 100') is 207,000 ft3 per Reference 4.10. The TSC envelope consists of all the areas depicted on the basement floor plan (Reference 4.10.a) excluding the south stairwell and area way open to above (southwest corner). The area between column lines B and G.2 and 2 and the fire wall running north-south is approximately 80' by 110' or 8,800 ft2. The area between column lines B and V and the fire wall runnin north-south and 5 is approximately 36' by (200' - 12') or 6,768 ft2. The total area is 8,800 ft + 6,768 ft2 or 15,568 ft2.

The height of the TSC envelope is determined per Reference 4.10.b, Cross Section 3. The basement and first floor finished elevations of 100' and 120' are used. A first floor thickness of 1' is assumed. The total height is (120' - 1') - 100' or 19'. The total volume of the TSC envelope is 15,568 ft2

19' or 295,792 ft3. It is assumed that 30 percent of the volume consists of components and/or structures, thus the total available free air volume is:

VTSC Envelope = 295,792 ft3

0.7 or 2.07E+05 ft3 For this accident analysis, it is assumed that manual operator action is required for the TSC to function in the emergency mode of operation. Consequently, it is conservatively assumed that the TSC never enters the emergency mode of operation (no filtration of make-up air through through the filtered air handling unit FFU-1.

TSC Normal Air Handling System flow rates, flow path and potential unfiltered leakage are determined below.

The Basement plan HVAC ductwork is provided in Reference 4.10.c. During normal operation, supply air enters the TSC envelope through the fan filter unit (FFU-1). The filters are by-passed and the make-up air is distributed throughout the TSC envelope with air handling unit AHU-0-1. Per Reference 4.10.d, AHU-0-1 has a capacity of 12,000 cfm.

D000040-119, Revision 0 PAGE 14 OF 70 Per Reference 4.11, Section 15940 - HVAC sequence of operation is as follows:

3.09 Technical Support Center (TSC)

A.

Non-Emergency Mode of Operation.

1.

AHU-0-1 shall operate as specified for VAV air handling unit(s) with static pressure optimization.

2.

Modulate return air damper in TSC space to maintain 1/8" wg positive pressure between the TSC space and the adjacent corridor.

Flow Rates under TSC Normal Operation Makeup air through AHU-0-1: 12,000 cfm It is assumed that the TSC does not isolate for the duration of the accident. A total of 13,000 cfm of unfiltered outside air is assumed to flow into and out of the TSC. 13,000 cfm conservatively bounds the expected design flow rate (AHU-0-1) of 12,000 cfm. The doses are conservatively calculated assuming no credit for filtration for the duration of the accident.

5.11 Reference 4.6 provides post-accident offsite x/Q's (sec/m3) at the EAB and LPZ. These values were utilized in other calculations to determine the offsite doses as a result of FSAR Chapter 15 design basis accidents. These X/Q's are included in the RADTRAD files as a "check" for model correctness.

Reference 4.7, Appendix 2 provides the post-accident X/Q's at the TSC. TSC atmospheric dispersion factors are not adjusted to reflect the anticipated post-accident occupancy factors of Reference 4.3.

Results are summarized in Table 4 below.

Table 4 TSC & Offsite Atmospheric Dispersion Factors Time Period EAB LPZ TSC 0 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 1.24E-04 3.9E-05 0 - 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 2.42E-05 2 - 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 3.3E-05 8 - 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 1.68E-05 1.6E-05 1 - 4 days 7.55E-06 1.2E-05 4 - 30 days 2.40E-06 8.7E-06

DC00040-119, Revision 0 PAGE 15 OF 70 5.12 Offsite and TSC breathing rates are per Reference 4.3, Sections 4.1.3, and 4.2.6.

Table 5 TSC& Offsite Breathing Rates Offsite EAB and LPZ)

TSC Time Rate (m"/sec)

Time Rate (m"/sec) 0-8 hr 3.5E-04 0-30 days 3.5E-04 8-24 hr 1.8E-04 1-30 days 2.3E-04 5.13 Per Reference 4.3, Section 4.2.6, TSC Occupancy Factors are as follows:

Table 6 TSC Occupancy Factors Time Occupancy Factor 0-24 hours 1.0 1-4 days 0.6 4-30 days 0.4 5.14 Isotopic decay constants are taken from FSAR Table 15A-2 and Reference 4.15.

Table 7 Isotopic Decay Constants Isotope Decay Constant 1/sec Isotope Decay Constant (1/sec)

Kr-85 2.04E-09 Xe-133 1.52E-06 Kr-85m 4.41E-05 Xe-135 2.11E-05 Kr-87 1.48E-04 Cs-134 1.07E-08 Kr-88 6.95E-05 Cs-136 6.10E-07 Rb-86 4.30E-07 Cs-137 7.28E-10 I-131 9.96E-07 1-132 8.26E-05 1-133 9.20E-06 1-134 2.20E-04 1-135 2.86E-05 (1) Rb-86 half-life = 18.66 days, Cs-134 half-life = 2.062 years (1) Cs-136 half-life = 13.16 days, Cs-137 half-life = 30.17 years

D000040-119, Revision 0 PAGE 16 OF 70 6.0 METHODOLOGY Consistent with the current licensing basis, as discussed in the VCS UFSAR Section 15.4.4 and RG 1.183, Appendix G (Reference 4.3), this analysis considers a conservative event with 1 percent fuel defects and 15 percent failed fuel.

Utilizing the assumptions in Section 2.0 and the design input data presented in Section 5.0, the Reactor Coolant Pump Locked Rotor accident, is calculated in compliance with RG 1.183 (Reference 4.3). This evaluation is provided in Section 7.0.

The applicable design inputs, as previously presented, are used to model the Reactor Coolant Pump Locked Rotor accident utilizing the RADTRAD computer code (Reference 4.4). The detailed description of the RADTRAD inputs can be found in Section 7.0. Attachment 2 provides the activity available for release to the environment from the breached fuel and existing equilibrium primary and secondary radioisotopic concentrations.

The conservative activity released to the environment is input into a NIF file (Attachment 3).

Based on assumption 2.17, the RFT file release time is set from 0 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . The release fractions for the 0 - 2, 2 - 8, and 8 - 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months time periods are based on the fraction of the activity released to the environment for a specified time period as a function of the total release of activity to the environment. The dose conversion factor INP file is the RADTRAD default file. The RADTRAD Input/Output File is I GPM-Conservative Case.oO.

DC00040-119, Revision 0 PAGE 17 OF 70 7.0 COMPUTATIONS 7.1 Conservative Case 7.1.1 Calculation of Activity Released to Environment Activity (concentration based on equilibrium reactor coolant 1% fuel defects and accident 15% failed fuel) released to the environment results from 2 sources; (a) activity from the primary system which leaks through the steam generator tubes and is released through the power relief valves and (b) and activity released from the secondary coolant.

A.

Primary system reactor coolant activity release to the environment is calculated as follows:

The equilibrium activity in the reactor core, based on 15% failed fuel is given in Section 5.2, Table 1.

The activity in the reactor core released from the gap, based on 15% failed fuel is calculated as follows.

Cp(ff) = Act

Frp *%FF

FG where:

Cp(ff)

=

activity of isotope i in the primary coolant in curies resulting from 15%

failed fuel Act curies of isotope i per core, including the uncertainty factor of 1.02 per Section 5.2, Table I Fan

=

maximum radial peaking factor assumed for damaged fuel assembly per Section 5.4 (unit less) 1.7

%FF

=

percent failed fuel per Section 5.4 (unit less)

=

15 percent FG fraction of isotope i activity in damaged fuel rods that escapes as a gap release per Section 5.4 (unit less).

Using this methodology, the design basis activity released to the primary coolant as a result of 15% failed fuel is provided in Attachment 2.

D000040-119, Revision 0 PAGE 18 OF 70 The equilibrium activity in the reactor coolant, based on 1% fuel defects, is provided in Section 5.3, Table 2 in uCi/gm. Assuming 1 pound is equal to 454 grams and 1 curie is equal to 1E+06 uCi, the equilibrium reactor coolant activity is converted to Ci The activity in the reactor coolant is calculated as follows.

Cp(fd) = Ac;

454* 1/IE+06

M where:

Cp(fd)

= equilibrium activity of isotope i in the primary coolant in curies resulting from 1% fuel defects Acs uCi/gm of isotope i in the primary equilibrium reactor coolant (based on 1% fuel defects) per Section 5.3, Table 2 M

=

mass of primary reactor coolant per Section 5.7 is 4.00E+05 lbs The activity released from the primary reactor coolant to the environment is calculated as follows:

Activity Released to Environment = Cp

1/M

Lp

PF

T Where:

Cp =

primary coolant concentration in Ci Cp =

Cp(ff) + Cp(fd)

Calculation of Cp is as follows:

Lp =

primary to secondary leak rate, 1.0 gpm 1.0 gpm

=

I gal/min

60 min/hr

1 ft3/7.4805 gal

62.4 lb/ft3 500.5013 lbs/hr The density of 62.4 lb/ft3 for the pressurized primary fluid for the primary to secondary leak rate is based on the guidance of Section 2.5 and utilized in the secondary coolant system equilibrium analysis (Reference 4.13).

PF =

steam generator partition coefficient is 0.01 for iodines and alkali metals; no holdup for the noble gases (1), Section 2.9 T = time in hours Using this methodology, the design basis activity from the primary reactor coolant (based on 15% failed fuel and I% fuel defects) available for release to the environment is provided in for the time periods 0 - 2, 2 - 8, and 8 - 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at 1.0 gpm primary to secondary system leak rates.

DC00040-119, Revision 0 PAGE 19 OF 70 B.

Secondary system activity is calculated utilizing the single member decay chain methodology of Reference 4.13, Section 1.2 assuming 1 pound is equal to 454 grams and 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months is equal to 3600 seconds as follows:

Csg = Lp

Cp

454/(Lb +?,

Msg

3600)

Csg = equilibrium activity concentration in secondary coolant for isotope i (uCi/lb)

Lp

= primary to secondary system leak rate (lb/hr). 1.0 gpm for the conservative case - Reference 4.13, Section 2.1(E):

1.0 gpm = 500.501303 lb/hr Cp = primary coolant activity concentration I% fuel defects (uCi/gm)

Lb

= steam generator blowdown (lbs/hr)

Conservative SG Blowdown = 12,756 lb/hr (Reference 4.17, Section 3.1.1) a,

= isotope decay constant (1 /sec), Section 5.14, Table 7 Msg = steam generator water mass (total of all 3 in lbs).

340,000 (Reference 4.14, Table 7-1)

Using this methodology, the design basis activity from the equilibrium secondary reactor coolant available for release is provided in Attachment 2.

D000040-119, Revision 0 PAGE 20 OF 70 The activity released from the equilibrium secondary reactor coolant to the environment is calculated as follows:

Activity Released = Cp

M / PF

IE+06 Cp

= secondary equilibrium coolant concentration in uCi/lb M

=

mass of the steam release from the 3 steam generators. Per Section 5.5, 447,900 lbs (0 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ), 868,300 lbs (2 - 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ), and 1,200,000 lbs (8 - 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ) 1/1E+06

=

conversion factor, uCi to Ci PF =

steam generator partition coefficient is 0.01 for iodines and alkali metals; no holdup for the noble gases (1)

Using this methodology, the design basis activity from the equilibrium secondary reactor coolant (based on 1% fuel defects) released to the environment is provided in Attachment 2 for the time periods 0 - 2, 2 - 8, and 8 - 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at 1.0 gpm primary to secondary system leak rates.

The total release from the steam generators to the environment is the sum of the primary and secondary releases for the 0 - 2, 2 - 8, and 8 - 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months time periods at 1.0 gpm primary to secondary system leak rates. Total releases as a function of time are presented in Attachment 2.

Computer program RADTRAD source input file (NIF file) is created based on the released activity calculated above and provided in Attachment 3. One file is created; a 0 - 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months release for the I gpm primary to secondary leak rates. The release fractions for the 0 - 2, 2 -

8, and 8 - 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months time periods are based on the fraction of the activity released to the environment for a specified time period as a function of the total release of activity to the environment. The associated release fraction file (RFT file) is provided in Attachment 4.

DO00040-119, Revision 0 PAGE 21 OF 70 7.1.2 RADTRAD Input Description A visual representation of the Locked Rotor event is given in Attachment 1 and developed below (per Sections 2.0, 5.0, 6.0, and 7.0).

The activity released to the environment is assumed to be released as a ground level release from the SG atmospheric relief valves or safety valves in a time period not to exceed 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (Section 2.14). All leakage (steaming through either the steam generator atmospheric relief valves or safety valves) is immediately released to the environment without holdup, plateout, filtration, or dilution.

The RADTRAD model input consists of 3 volumes, 4 flow pathways, and 3 dose locations.

Source inventory (activity available for release to the environment) is input as user defined NIF file provided in Attachment 3 for the 0 - 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months time period for the 1.0 gpm primary to secondary reactor coolant leak rate.

Release fractions and timing data for the Locked Rotor event are input as a user defined RFT files provided in Attachment 4 for the 0 - 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months time period.

7.1.2.1 RADTRAD Volume 1 represents the activity available for release to the environment:

Volume is arbitrarily modeled as 1E+04 ft3 (this has no impact on results) 7.1.2.2 RADTRAD Volume 2 represents the Environment:

No input required 7.1.2.3 RADTRAD Volume 3 represents the TSC:

TSC habitability volume equals 2.07E+05 ft3 (Section 5.10) 7.1.2.4 RADTRAD Pathway 1 represents the activity release to the environs:

Release rate is arbitrarily modeled as 1E+10 cfm to ensure a complete release of the source (this is conservative). No removal mechanisms, such as filtration or holdup are assumed to reduce the source term.

7.1.2.5 RADTRAD Pathway 2 represents the TSC Outside Air Makeup Pathway:

Normal Mode - Outside air intake equals 12,000 cfm 0 minutes through the duration of the accident No Emergency Mode or Filtration 7.1.2.6 RADTRAD Pathway 3 represents the TSC Unfiltered Air Inleakage Pathway:

TSC unfiltered air inleakage is conservatively assumed to be 1,000 cfm 0 minutes through the duration of the accident No filter is applied to this pathway 7.1.2.7 RADTRAD Pathway 4 represents the TSC Exhaust Air Pathway:

Exhaust air flow equals intake and inleakage, 12,000 cfm plus 1,000 cfm =

minutes through the duration of the accident) 13,000 cfm 0 No filter is applied to this pathway

DO00040-119, Revision 0 PAGE 22 OF 70 7.1.2.8 RADTRAD Dose Location 1 - EAB:

X/Q value per Table 4used for 0 - 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months Breathing rate values per Table No. 6 (0 - 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months ) 7.1.2.9 RADTRAD Dose Location 2 - LPZ:

X/Q values per Table 4 Breathing rate values per Table 7 7.1.2.10 RADTRAD Dose Location 3-Protected TSC:

X/Q values per Table 4 Breathing rate per Table 5 Occupancy values per Table 6 7.1.2.11 RADTRAD Source Term:

Use the user defined source inventory file 0 - 24 Hr Design.nif for the I gpm primary to secondary leak rate. This file is determined in Attachment No. 2 and provided in.

Modeled power level as 1.0453 MWIh, to obtain total I-131 activity release at 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , 1.0443 for increased inleakage case.

Model isotopic decay and daughter in-growth Use the user defined RADTRAD release fraction files, 0 - 24 Hr Design.rft.

Use the default RADTRAD FGR 11 & 12 dose conversion factors for the MACCS 60 isotope inventory, Fgrl l &12.inp 7.1.2.12 RADTRAD Control Options:

Applicable Control Options are selected for the additional data supplied in the output printout The resulting RADTRAD output file is: 1 GPM Conservative Case.oO (Attachment 5)

D000040-119, Revision 0 PAGE 23 OF 70 8.0

SUMMARY

OF RESULTS Table No. 8: RCPLR Dose in Rems TEDE - TSC Case Inte grated Dose Rem - TEDE Attachment EAB LPZ TSC NRC Acceptance Criteria 2.5 2.5 5.0 RCPLR 0.67 0.66 0.83 5

Per Section 1.3, the dose acceptance criterion (References 4.1 & 4.3) for the Reactor Coolant Pump Locked Rotor event is given as 5 Rem TEDE for the TSC. Based on the results in Table 8, the TSC dose resulting from the Reactor Coolant Pump Locked Rotor event is acceptable.

The EAB and LPZ doses are the same as those provided in the AST RCPLR calculation D000040-100, Revision 1 (Reference 4.19).

9.0 DISPOSITION OF RESULTS The results of this calculation are incorporated into the VCSNS FSAR 15.4.4.

DO00040-119, Revision 0 Page 24 of 70 RADTRAD Model

D000040-119, Revision 0 Page 25 of 70 Volume I Source Volume 1.0+04 ft3 (arbitrary volume)

FP 1 1.0E+10 cfm (arbitrary flow rate)

No filtration or holdup RADTRAD Locked Rotor Model Volume 2 Environment FP 2 - 12,000 cfm unfiltered makeup (0 - 720 hr.

FP 3 - 1,000 cfm unfiltered inleakage (0 - 720 hr.

Volume 3 TSC Habitability Envelope (2.07E+05 ft3)

FP 4 - TSC exhaust: makeup plus unfiltered inleakage 13,000 cfm (0 - 720 hr.)

DO00040-119, Revision 0 Page 26 of 70 Conservative Source Terms

DO00040-119, Revision 0 Page 27 of 70 Primary Reactor Coolant - Conservative Case Calculation of Activity Released to the Primary Reactor Coolant

- 15% Failed Fuel from the Gap (Conservative Case Only)

Isotopes Core(')

Activity Ci Radial Peaking Factor(2)

Percent Failed Fuel (2, Available Gap Activity Ci Gap Release Fraction (2)

Primary RC Activity Ci Kr-85 8.30E+05 1.7 0.15 2.12E+05 0.2 4.23E+04 Kr-85m 2.72E+07 1.7 0.15 6.94E+06 0.1 6.94E+05 Kr-87 4.96E+07 1.7 0.15 1.26E+07 0.1 1.26E+06 Kr-88 6.71 E+07 1.7 0.15 1.71E+07 0.1 1.71E+06 Rb-86 4.43E+04 1.7 0.15 1.13E+04 0.24 2.71E+03 1-131 8.20E+07 1.7 0.15 2.09E+07 0.16 3.35E+06 1-132 1.20E+08 1.7 0.15 3.06E+07 0.1 3.06E+06 1-133 1.68E+08 1.7 0.15 4.28E+07 0.1 4.28E+06 1-134 1.80E+08 1.7 0.15 4.59E+07 0.1 4.59E+06 1-135 1.54E+08 1.7 0.15 3.93E+07 0.1 3.93E+06 Xe-133 1.70E+08 1.7 0.15 4.34E+07 0.1 4.34E+06 Xe-135 3.70E+07 1.7 0.15 9.44E+06 0.1 9.44E+05 Cs-134 1.01E+07 1.7 0.15 2.58E+06 0.24 6.18E+05 Cs-136 3.08E+06 1.7 0.15 7.85E+05 0.24 1.88E+05 Cs-137 5.66E+06 1.7 0.15 1.44E+06 0.24 3.46E+05 Notes:

1. Section 5.2, Table 1

2. Section 5.4

D000040-119, Revision 0 Page 28 of 70 Primary Reactor Coolant - Conservative Case Calculation of Total Primary Coolant Activity Based on 1 % Fuel Defects sotopes 1% FD Primary Coolant Concentratio&11 uCi/gm Primary Coolant Concentration (2)

Ci/Ib Primary Coolant Mass (3)

Ibm Primary Coolant Activity Ci Total Activity Released to Coolant(4)

Ci Kr-85 7.60E+00 3.45E-03 4.0E+05 1.38E+03 4.37E+04 Kr-85m 1.80E+00 8.17E-04 4.0E+05 3.27E+02 6.94E+05 Kr-87 1.10E+00 4.99E-04 4.0E+05 2.00E+02 1.26E+06 Kr-88 3.20E+00 1.45E-03 4.0E+05 5.81E+02 1.71E+06 Rb-86 3.60E-02 1.63E-05 4.0E+05 6.54E+00 2.72E+03 1-131 3.00E+00 1.36E-03 4.0E+05 5.45E+02 3.35E+06 1-132 3.10E+00 1.41 E-03 4.0E+05 5.63E+02 3.06E+06 1-133 4.60E+00 2.09E-03 4.0E+05 8.35E+02 4.28E+06 1-134 6.00E-01 2.72E-04 4.0E+05 1.09E+02 4.59E+06 1-135 2.40E+00 1.09E-03 4.0E+05 4.36E+02 3.93E+06 Xe-133 2.90E+02 1.32E-01 4.0E+05 5.27E+04 4.39E+06 Xe-135 8.60E+00 3.90E-03 4.0E+05 1.56E+03 9.45E+05 Cs-134 4.40E+00 2.00E-03 4.0E+05 7.99E+02 6.19E+05 Cs-136 4.50E+00 2.04E-03 4.0E+05 8.17E+02 1.89E+05 Cs-137 2.10E+00 9.53E-04 4.0E+05 3.81E+02 3.47E+05 Notes:

1. Section 5.3, Table 2, based on 1 % fuel defect

2. Conversion Factor = 454 gm/lb

1 Ci/1 E+06 uCi

3. Section 5.7

4. 15% failed fuel + 1 % fuel defect

D000040-119, Revision 0 Page 29 of 70 Primary Reactor Coolant - Conservative Case Calculation of Activity Release to the Environment for Primary Reactor Coolant in the Steam Generator During This Event - 1.0 GPM Leakage sotopes Total Coolant Activity Ci Primary Coolant Mass')

lbs Primary Coolant Concentration Ci/Ib Leak Rate (2) lb/hr Activity Release Ci/hr Kr-85 4.37E+04 4.0E+05 1.09E-01 500.5013 5.47E+01 Kr-85m 6.94E+05 4.0E+05 1.73E+00 500.5013 8.68E+02 Kr-87 1.26E+06 4.0E+05 3.16E+00 500.5013 1.58E+03 Kr-88 1.71E+06 4.0E+05 4.28E+00 500.5013 2.14E+03 Rb-86 2.72E+03 4.0E+05 6.79E-03 500.5013 3.40E+00 1-131 3.35E+06 4.0E+05 8.37E+00 500.5013 4.19E+03 1-132 3.06E+06 4.0E+05 7.65E+00 500.5013 3.83E+03 1-133 4.28E+06 4.0E+05 1.07E+01 500.5013 5.36E+03 I-134 4.59E+06 4.0E+05 1.15E+01 500.5013 5.74E+03 1-135 3.93E+06 4.0E+05 9.82E+00 500.5013 4.91 E+03 Xe-133 4.39E+06 4.0E+05 1.10E+01 500.5013 5.49E+03 Xe-135 9.45E+05 4.0E+05 2.36E+00 500.5013 1.18E+03 Cs-134 6.19E+05 4.0E+05 1.55E+00 500.5013 7.74E+02 Cs-136 1.89E+05 4.0E+05 4.73E-01 500.5013 2.37E+02 Cs-137 3.47E+05 4.0E+05 8.67E-01 500.5013 4.34E+02 Notes:

1. Section 5.7

2. Section 7.1.1

D000040-119, Revision 0 Page 30 of 70 Primary Reactor Coolant - Conservative Case Calculation of Activity Released as a Function of Time to the Environment from Primary Coolant in the Steam Generator - 1.0 GPM Leakage sotopes Activity Released Rate Ci/Hr Partition Fraction(i)

Activity Released 0 - 2 Hrs.

Ci Activity Not Released 0 - 2 Hrs.

Ci2 Activity Released 2 - 8 Hrs.

Ci Activity Not Released 2 - 8 Hrs.

Ci2 Activity Released 8 - 24 Hrs.

Ci Activity Not Released 8 - 24 Hrs.

Ci2 Kr-85 5.47E+01 1

1.09E+02 0.00E+00 3.28E+02 0.00E+00 8.75E+02 0.00E+00 Kr-85m 8.68E+02 1

1.74E+03 0.00E+00 5.21E+03 0.00E+00 1.39E+04 0.00E+00 Kr-87 1.58E+03 1

3.17E+03 0.00E+00 9.50E+03 0.00E+00 2.53E+04 0.00E+00 Kr-88 2.14E+03 1

4.28E+03 0.00E+00 1.29E+04 0.00E+00 3.43E+04 0.00E+00 Rb-86 3.40E+00 0.01 6.80E-02 6.73E+00 2.04E-01 2.02E+01 5.44E-01 5.39E+01 I-131 4.19E+03 0.01 8.37E+01 8.29E+03 2.51E+02 2.49E+04 6.70E+02 6.63E+04 1-132 3.83E+03 0.01 7.66E+01 7.58E+03 2.30E+02 2.27E+04 6.13E+02 6.07E+04 I-133 5.36E+03 0.01 1.07E+02 1.06E+04 3.22E+02 3.18E+04 8.58E+02 8.49E+04 I-134 5.74E+03 0.01 1.15E+02 1.14E+04 3.45E+02 3.41E+04 9.19E+02 9.10E+04 I-135 4.91E+03 0.01 9.83E+01 9.73E+03 2.95E+02 2.92E+04 7.86E+02 7.78E+04 Xe-133 5.49E+03 1

1.10E+04 0.00E+00 3.29E+04 0.00E+00 8.78E+04 0.00E+00 Xe-135 1.18E+03 1

2.37E+03 0.00E+00 7.10E+03 0.00E+00 1.89E+04 0.00E+00 Cs-134 7.74E+02 0.01 1.55E+01 1.53E+03 4.65E+01 4.60E+03 1.24E+02 1.23E+04 Cs-136 2.37E+02 0.01 4.74E+00 4.69E+02 1.42E+01 1.41E+03 3.79E+01 3.75E+03 Cs-137 4.34E+02 0.01 8.68E+00 8.59E+02 2.60E+01 2.58E+03 6.94E+01 6.87E+03 Notes:

1. Section 2.9

2. Activity not released is equal to the difference between the activity released for a given time period with and without accounting for the partition factor.

D000040-119, Revision 0 Page 31 of 70 Secondary Reactor Coolant - Conservative Cases Secondary Coolant Specific Activity based on Reference 4.14. Table 5-15 LB')

Conservative SG Blowdown =

12756 Ibm/hr MSG (1)

Water Mass in All 3 SGs =

340000 Ibm LP(1 )

Leak Rate =

1 gpm =

500.501303 Ibm/hr Decay (2)

RC (3)

Sec. Coolant Eq. Conc.

Csg uCi/lb 4 5 Isotopes Constant 1/sec Concentration Ci/ m 1

m Rb-86 4.30E-07 3.60E-02 6.16E-01 1-131 9.96E-07 3.00E+00 4.88E+01 1-132 8.26E-05 3.10E+00 6.19E+00 1-133 9.20E-06 4.60E+00 4.35E+01 1-134 2.20E-04 6.00E-01 4.83E-01 1-135 2.86E-05 2.40E+00 1.14E+01 Cs-134 1.07E-08 4.40E+00 7.83E+01 Cs-136 6.10E-07 4.50E+00 7.57E+01 Cs-137 7.28E-10 2.10E+00 3.74E+01 Notes:

1. Sections 7.1.1(B) & 7.2.1(B)

2. Section 5.14, Table 7

3. Reference 4.14, Table 5-15

4. RC concentrations based on 1 % failed fuel. Conservative case assumes

5. Methodology per Sections 7.1.1 and 7.2.1 B

D000040-119, Revision 0 Page 32 of 70 Secondary Reactor Coolant - Conservative Case Calculation of Activity Released as a Function of Time to the Environment from Secondary Coolant in the Steam Generator - 1.0 GPM Leakage (0 - 2 Hrs.)

sotopes Secondary Equilibrium Concentration uCi/Ib Mass Release Rate 0 - 2 Hrs.

Ibm(')

Activity Available for Release 0 - 2 Hrs.

Ci I GPM Leakage Activity Not Released to Environment 0 - 2 Hrs.

Ci Total Activity Available for Release 0 - 2 Hrs.

Ci artition Factor(2)

Activity Released to Environment 0 - 2 Hrs.

Ci Rb-86 6.16E-01 447900 2.76E-01 6.73E+00 7.01E+00 100 7.01 E-02 1-131 4.88E+01 447900 2.18E+01 8.29E+03 8.31E+03 100 8.31 E+01 1-132 6.19E+00 447900 2.77E+00 7.58E+03 7.59E+03 100 7.59E+01 1-133 4.35E+01 447900 1.95E+01 1.06E+04 1.06E+04 100 1.06E+02 1-134 4.83E-01 447900 2.17E-01 1.14E+04 1.14E+04 100 1.14E+02 1-135 1.14E+01 447900 5.11E+00 9.73E+03 9.74E+03 100 9.74E+01 Cs-134 7.83E+01 447900 3.51E+01 1.53E+03 1.57E+03 100 1.57E+01 Cs-136 7.57E+01 447900 3.39E+01 4.69E+02 5.03E+02 100 5.03E+00 Cs-137 3.74E+01 447900 1.68E+01 8.59E+02 8.76E+02 100 8.76E+00 Notes:

1. Section 5.5

2. Section 2.9

D000040-119, Revision 0 Page 33 of 70 Secondary Reactor Coolant - Conservative Case Calculation of Activity Released as a Function of Time to the Environment from Secondary Coolant in the Steam Generator - 1.0 GPM Leakage (2 - 8 Hrs.)

sotopes Secondary Equilibrium Concentration uCi/Ib Mass Release Rate 2 - 8 Hrs.

Ibm(')

Activity Available for Release 2 - 8 Hrs.

Ci I GPM Leakage Activity Not Released to Environment 2 - 8 Hrs.

Ci Total Activity Available for Release 2 - 8 Hrs.

Ci artition Factor (2)

Activity Released to Environment 2 - 8 Hrs.

Ci Rb-86 6.16E-01 868300 5.35E-01 2.02E+01 2.07E+01 100 2.07E-01 1-131 4.88E+01 868300 4.24E+01 2.49E+04 2.49E+04 100 2.49E+02 1-132 6.19E+00 868300 5.37E+00 2.27E+04 2.28E+04 100 2.28E+02 1-133 4.35E+01 868300 3.78E+01 3.18E+04 3.19E+04 100 3.19E+02 1-134 4.83E-01 868300 4.20E-01 3.41E+04 3.41E+04 100 3.41E+02 1-135 1.14E+01 868300 9.91E+00 2.92E+04 2.92E+04 100 2.92E+02 Cs-134 7.83E+01 868300 6.80E+01 4.60E+03 4.67E+03 100 4.67E+01 Cs-136 7.57E+01 868300 6.58E+01 1.41E+03 1.47E+03 100 1.47E+01 Cs-137 3.74E+01 868300 3.25E+01 2.58E+03 2.61E+03 100 2.61E+01 Notes:

1. Section 5.5

2. Section 2.9

D000040-119, Revision 0 Page 34 of 70 Secondary Reactor Coolant - Conservative Case Calculation of Activity Released as a Function of Time to the Environment from Secondary Coolant in the Steam Generator - 1.0 GPM Leakage (8 - 24 Hrs.)

sotopes Secondary Equilibrium Concentration uCi/Ib Mass Release Rate 8 - 24 Hrs.

lbm0)

Activity Available for Release 8 - 24 Hrs.

Ci 1 GPM Leakage Activity Not Released to Environment 8 - 24 Hrs.

Ci Total Activity Available for Release 8 - 24 Hrs.

Ci artition Factor(2)

Activity Released to Environment 8 - 24 Hrs.

Ci Rb-86 6.16E-01 1.20E+06 7.39E-01 5.39E+01 5.46E+01 100 5.46E-01 1-131 4.88E+01 1.20E+06 5.85E+01 6.63E+04 6.64E+04 100 6.64E+02 1-132 6.19E+00 1.20E+06 7.42E+00 6.07E+04 6.07E+04 100 6.07E+02 1-133 4.35E+01 1.20E+06 5.22E+01 8.49E+04 8.50E+04 100 8.50E+02 1-134 4.83E-01 1.20E+06 5.80E-01 9.10E+04 9.10E+04 100 9.10E+02 1-135 1.14E+01 1.20E+06 1.37E+01 7.78E+04 7.79E+04 100 7.79E+02 Cs-134 7.83E+01 1.20E+06 9.40E+01 1.23E+04 1.24E+04 100 1.24E+02 Cs-136 7.57E+01 1.20E+06 9.09E+01 3.75E+03 3.84E+03 100 3.84E+01 Cs-137 3.74E+01 1.20E+06 4.49E+01 6.87E+03 6.92E+03 100 6.92E+01 Notes:

1. Section 5.5

2. Section 2.9

D000040-119, Revision 0 Page 35 of 70 Primary & Secondary Reactor Coolant Release - Conservative Case Calculation of Total Activity (Primary & Secondary) Released as a Function of Time to the Environment from the Steam Generators - 1.0 GPM Leakage uclide Primary Coolant Activity Release 0 - 2 Hrs.

Ci Secondary Coolant Activity Release 0 - 2 Hrs.

Ci Total Activity Release 0 - 2 Hrs.

Ci Primary Coolant Activity Release 2 - 8 Hrs.

Ci Secondary Coolant Activity Release 2 - 8 Hrs.

Ci Total Activity Release 2 - 8 Hrs.

Ci Primary Coolant Activity Release 8 - 24 Hrs.

Ci Secondary Coolant Activity Release 8 - 24 Hrs.

Ci Total Activity Release 8 - 24 Hrs.

Ci Total Activity Release 0 - 24 Hrs.

Ci Kr-85 1.09E+02 0.00E+00 1.09E+02 3.28E+02 0.00E+00 3.28E+02 8.75E+02 0.00E+00 8.75E+02 1.31E+03 Kr-85m 1.74E+03 0.00E+00 1.74E+03 5.21E+03 0.00E+00 5.21E+03 1.39E+04 0.00E+00 1.39E+04 2.08E+04 Kr-87 3.17E+03 0.00E+00 3.17E+03 9.50E+03 0.00E+00 9.50E+03 2.53E+04 0.00E+00 2.53E+04 3.80E+04 Kr-88 4.28E+03 0.00E+00 4.28E+03 1.29E+04 0.00E+00 1.29E+04 3.43E+04 0.00E+00 3.43E+04 5.14E+04 Rb-86 6.80E-02 7.01 E-02 1.38E-01 2.04E-01 2.07E-01 4.11 E-01 5.44E-01 5.46E-01 1.09E+00 1.64E+00 1-131 8.37E+01 8.31E+01 1.67E+02 2.51E+02 2.49E+02 5.00E+02 6.70E+02 6.64E+02 1.33E+03 2.00E+03 1-132 7.66E+01 7.59E+01 1.52E+02 2.30E+02 2.28E+02 4.57E+02 6.13E+02 6.07E+02 1.22E+03 1.83E+03 1-133 1.07E+02 1.06E+02 2.14E+02 3.22E+02 3.19E+02 6.41E+02 8.58E+02 8.50E+02 1.71E+03 2.56E+03 1-134 1.15E+02 1.14E+02 2.29E+02 3.45E+02 3.41E+02 6.86E+02 9.19E+02 9.10E+02 1.83E+03 2.74E+03 1-135 9.83E+01 9.74E+01 1.96E+02 2.95E+02 2.92E+02 5.87E+02 7.86E+02 7.79E+02 1.56E+03 2.35E+03 Xe-133 1.10E+04 0.00E+00 1.10E+04 3.29E+04 0.00E+00 3.29E+04 8.78E+04 0.00E+00 8.78E+04 1.32E+05 Xe-135 2.37E+03 0.00E+00 2.37E+03 7.10E+03 0.00E+00 7.10E+03 1.89E+04 0.00E+00 1.89E+04 2.84E+04 Cs-134 1.55E+01 1.57E+01 3.12E+01 4.65E+01 4.67E+01 9.31E+01 1.24E+02 1.24E+02 2.48E+02 3.72E+02 Cs-136 4.74E+00 5.03E+00 9.77E+00 1.42E+01 1.47E+01 2.89E+01 3.79E+01 3.84E+01 7.63E+01 1.15E+02 Cs-137 8.68E+00 8.76E+00 1.74E+01 2.60E+01 2.61E+01 5.21E+01 6.94E+01 6.92E+01 1.39E+02 2.08E+02

D000040-119, Revision 0 Page 36 of 70 Primary & Secondary Reactor Coolant Release Fractions - Conservative Case Release Fractions as a Function of Time for RADTRAD RFT File Input Nuclide 0-2Hrs.

Hours 2-8Hrs.

Hours 8-24Hrs.

Hours Kr-85 8.33E-02 2.50E-01 6.67E-01 Kr-85m 8.33E-02 2.50E-01 6.67E-01 Kr-87 8.33E-02 2.50E-01 6.67E-01 Kr-88 8.33E-02 2.50E-01 6.67E-01 Rb-86 8.42E-02 2.51 E-01 6.65E-01 1-131 8.34E-02 2.50E-01 6.67E-01 1-132 8.33E-02 2.50E-01 6.67E-01 1-133 8.34E-02 2.50E-01 6.67E-01 1-134 8.33E-02 2.50E-01 6.67E-01 1-135 8.33E-02 2.50E-01 6.67E-01 Xe-133 8.33E-02 2.50E-01 6.67E-01 Xe-135 8.33E-02 2.50E-01 6.67E-01 Cs-134 8.38E-02 2.51 E-01 6.66E-01 Cs-136 8.49E-02 2.52E-01 6.64E-01 Cs-137 8.38E-02 2.50E-01 6.66E-01

DO00040-119, Revision 0 Page 37 of 70 RADTRAD Source Input File for Conservative Case

D000040-119, Revision 0 Page 38 of 70 Nuclide Inventory Name:0-24 Design.nif Normalized MACCS Sample 3412 MWth PWR Core Inventory Power Level:

0. 1000E+01 Nuclides:

60 Nuclide 001:

Co-58 7

0.6117120000E+07 0.5800E+02 0.0000E+00 none 0.0000E+00 none 0.0000E+00 none 0.0000E+00 Nuclide 002:

Co-60 7

0.1663401096E+09 0.6000E+02 0.0000E+00 none 0.0000E+00 none 0.0000E+00 none 0.0000E+00 Nuclide 003:

Kr-85 1

0.3382974720E+09 0.8500E+02 1.31E+03 none 0.0000E+00 none 0.0000E+00 none 0.0000E+00 Nuclide Kr-85m 004:

1 0.1612800000E+05 0.8500E+02 2.08E+04 Kr-85 0.2100E+00 none 0.0000E+00 none 0.0000E+00 Nuclide 005:

Kr-87 1

0.4578000000E+04 0.8700E+02 3.80E+04 Rb-87 0.1000E+01 none 0.0000E+00 none 0.0000E+00 Nuclide 006:

Kr-88 1

D000040-119, Revision 0 Page 39 of 70 0.1022400000E+05 0.8800E+02 5.14E+04 Rb-88 0.1000E+01 0

none 0.0000E+00 none 0.0000E+00 Nuclide 007:

Rb-86 3

0.1612224000E+07 0.8600E+02 1.64E+00 none 0.0000E+00 none 0.0000E+00 none 0.0000E+00 Nuclide 008:

Sr-89 5

0.4363200000E+07 0.8900E+02 0.0000E+00 none 0.0000E+00 none 0.0000E+00 none 0.0000E+00 Nuclide 009:

Sr-90 5

0.9189573120E+09 0.9000E+02 0.0000E+00 Y-90 0.1000E+01 none 0.0000E+00 none 0.0000E+00 Nuclide 010:

Sr-91 5

0.3420000000E+05

0. 9100E+02 0.0000E+00 Y-91m 0.5800E+00 Y-91 0.4200E+00 none 0.0000E+00 Nuclide 011:

Sr-92 5

0.9756000000E+04 0.9200E+02 0.0000E+00 Y-92 0.1000E+01 none 0.0000E+00 none 0.0000E+00 Nuclide 012:

Y-90 9

0.2304000000E+06

0.9000E+02 0.0000E+00 none 0.0000E+00 none 0.0000E+00 none 0.0000E+00 Nuclide 013:

Y-91 9

0.5055264000E+07 0.9100E+02 0.0000E+00 none 0.0000E+00 none 0.0000E+00 none 0.0000E+00 Nuclide 014:

Y-92 9

0.1274400000E+05 0.9200E+02 0.0000E+00 none 0.0000E+00 none 0.0000E+00 none 0.0000E+00 Nuclide 015:

Y-93 9

0.3636000000E+05 0.9300E+02 0.0000E+00 Zr-93 0.1000E+01 none 0.0000E+00 none 0.0000E+00 Nuclide 016:

Zr-95 9

0.5527872000E+07 0.9500E+02 0.0000E+00 Nb-95m 0.7000E-02 Nb-95 0.9900E+00 none 0.0000E+00 Nuclide 017:

Zr-97 9

0.6084000000E+05 0.9700E+02 0.0000E+00 Nb-97m 0.9500E+00 Nb-97 0.5300E-01 none 0.0000E+00 Nuclide 018:

Nb-95 9

0.3036960000E+07 0.9500E+02

D000040-119, Revision 0 Page 41 of 70 0.0000E+00 none 0.0000E+00 none 0.0000E+00 none 0.0000E+00 Nuclide 019:

Mo-99 7

0.2376000000E+06 0.9900E+02 0.0000E+00 Tc-99m 0.8800E+00 Tc-99 0.1200E+00 none 0.0000E+00 Nuclide Tc-99m 020:

7 0.2167200000E+05 0.9900E+02 0.0000E+00 Tc-99 0.1000E+01 none 0.0000E+00 none 0.0000E+00 Nuclide 021:

Ru-103 7

0.3393792000E+07 0.1030E+03 0.0000E+00 Rh-103m 0.1000E+01 none 0.0000E+00 none 0.0000E+00 Nuclide 022:

Ru-105 7

0.1598400000E+05

0. 1050E+03 O.0000E+00 Rh-105 0.1000E+01 none 0.0000E+00 none 0.0000E+00 Nuclide 023:

Ru-106 7

0.3181248000E+08

0. 1060E+03 0.0000E+00 Rh-106 0.1000E+01 none 0.0000E+00 none 0.0000E+00 Nuclide 024:

Rh-105 7

0.1272960000E+06 0.1050E+03 0.0000E+00

D000040-119, Revision 0 Page 42 of 70 none 0.0000E+00 none 0.0000E+00 none 0.0000E+00 Nuclide 025:

Sb-127 4

0.3326400000E+06

0. 1270E+03 0.0000E+00 Te-127m 0.1800E+00 Te-127 0.8200E+00 none 0.0000E+00 Nuclide 026:

Sb-129 4

0.1555200000E+05 0.1290E+03 0.0000E+00 Te-129m 0.2200E+00 Te-129 0.7700E+00 none 0.0000E+00 Nuclide 027:

Te-127 4

0.3366000000E+05 0.1270E+03 0.0000E+00 none 0.0000E+00 none 0.0000E+00 none 0.0000E+00 Nuclide 028:

Te-127m 4

0.9417600000E+07 0.1270E+03 0.0000E+00 Te-127 0.9800E+00 none 0.0000E+00 none 0.0000E+00 Nuclide 029:

Te-129 4

0.4176000000E+04 0.1290E+03 0.0000E+00 1-129 0.1000E+01 none 0.0000E+00 none 0.0000E+00 Nuclide 030:

Te-129m 4

0.2903040000E+07 0.1290E+03 0.0000E+00 Te-129 0.6500E+00

D000040-119, Revision 0 Page 43 of 70 1-129 0.3500E+00 none 0.0000E+00 Nuclide 031:

Te-131m 4

0.1080000000E+06 0.1310E+03 0.0000E+00 Te-131 0.2200E+00 1-131 0.7800E+00 none 0.0000E+00 Nuclide 032:

Te-132 4

0.2815200000E+06 0.1320E+03 0.0000E+00 1-132 0.1000E+01 none 0.0000E+00 none 0.0000E+00 Nuclide 033:

1-131 2

0.6946560000E+06 0.1310E+03 2.00E+03 Xe-131m 0.1100E-01 none 0.0000E+00 none 0.0000E+00 Nuclide 034:

1-132 2

0.8280000000E+04 0.1320E+03 1.83E+03 none 0.0000E+00 none 0.0000E+00 none 0.0000E+00 Nuclide 035:

1-133 2

0.7488000000E+05 0.1330E+03 2.56E+03 Xe-133m 0.2900E-01 Xe-133 0.9700E+00 none 0.0000E+00 Nuclide 036:

1-134 2

0.3156000000E+04 0.1340E+03 2.74E+03 none 0.0000E+00 none 0.0000E+00

D000040-119, Revision 0 Page 44 of 70 none 0.0000E+00 Nuclide 037:

1-135 2

0.2379600000E+05 0.1350E+03 2.35E+03 Xe-135m 0.1500E+00 Xe-135 0.8500E+00 none 0.0000E+00 Nuclide 038:

Xe-133 1

0.4531680000E+06 0.1330E+03 1.32E+05 none 0.0000E+00 none 0.0000E+00 none 0.0000E+00 Nuclide 039:

Xe-135 1

0.3272400000E+05 0. 1350E+03 2.84E+04 Cs-135 0.1000E+01 none 0.0000E+00 none 0.0000E+00 Nuclide 040:

Cs-134 3

0.6507177120E+08 0.1340E+03

3. 72E+02 none 0.0000E+00 none 0.0000E+00 none 0.0000E+00 Nuclide 041:

Cs-136 3

0.1131840000E+07 0.1360E+03 1.15E+02 none 0.0000E+00 none 0.0000E+00 none 0.0000E+00 Nuclide 042:

Cs-137 3

0.9467280000E+09 0.1370E+03 2.08E+02 Ba-137m 0.9500E+00 none 0.0000E+00 none 0.0000E+00

D000040-119, Revision 0 Page 45 of 70 Nuclide 043:

Ba-139 6

0.4962000000E+04 0.1390E+03 0.0000E+00 none 0.0000E+00 none 0.0000E+00 none 0.0000E+00 Nuclide 044:

Ba-140 6

0.1100736000E+07 0.1400E+03 0.0000E+00 La-140 0.1000E+01 none 0.0000E+00 none 0.0000E+00 Nuclide 045:

La-140 9

0.1449792000E+06 0.1400E+03 0.0000E+00 none 0.0000E+00 none 0.0000E+00 none 0.0000E+00 Nuclide 046:

La-141 9

0.1414800000E+05 0.1410E+03 0.0000E+00 Ce-141 0.1000E+01 none 0.0000E+00 none 0.0000E+00 Nuclide 047:

La-142 9

0.5550000000E+04 0.1420E+03 0.0000E+00 none 0.0000E+00 none 0.0000E+00 none 0.0000E+00 Nuclide 048:

Ce-141 8

0.2808086400E+07 0.1410E+03 0.0000E+00 none 0.0000E+00 none 0.0000E+00 none 0.0000E+00 Nuclide 049:

D000040-119, Revision 0 Page 46 of 70 Ce-143 8

0.1188000000E+06

0. 1430E+03 0.0000E+00 Pr-143 0.1000E+01 none 0.0000E+00 none 0.0000E+00 Nuclide 050:

Ce-144 8

0.2456352000E+08 0.1440E+03 0.0000E+00 Pr-144m 0.1800E-01 Pr-144 0.9800E+00 none 0.0000E+00 Nuclide 051:

Pr-143 9

0.1171584000E+07 0.1430E+03 0.0000E+00 none 0.0000E+00 none 0.0000E+00 none 0.0000E+00 Nuclide 052:

Nd-147 9

0.9486720000E+06 0.1470E+03 0.0000E+00 Pm-147 0.1000E+01 none 0.0000E+00 none 0.0000E+00 Nuclide 053:

Np-239 8

0.2034720000E+06 0.2390E+03 0.0000E+00 Pu-239 0.1000E+01 none 0.0000E+00 none 0.0000E+00 Nuclide 054:

Pu-238 8

0.2768863824E+10 0.2380E+03 0.0000E+00 U-234 0.1000E+01 none 0.0000E+00 none 0.0000E+00 Nuclide 055:

Pu-239

D000040-119, Revision 0 Page 47 of 70 8

0.7594336440E+12 0.2390E+03 0.0000E+00 U-235 0.1000E+01 none 0.0000E+00 none 0.0000E+00 Nuclide 056:

Pu-240 8

0.2062920312E+12 0.2400E+03 0.0000E+00 U-236 0.1000E+01 none 0.0000E+00 none 0.0000E+00 Nuclide 057:

Pu-241 8

0.4544294400E+09

0. 2410E+03 0.0000E+00 U-237 0.2400E-04 Am-241 0.1000E+01 none 0.0000E+00 Nuclide 058:

Am-241 9

0.1363919472E+11 0.2410E+03 0.0000E+00 Np-237 0.1000E+01 none 0.0000E+00 none 0.0000E+00 Nuclide 059:

Cm-242 9

0.1406592000E+08 0.2420E+03 0.0000E+00 Pu-238 0.1000E+01 none 0.0000E+00 none 0.0000E+00 Nuclide 060:

Cm-244 9

0.571508136E+9 0.2440E+03 0.0000E+00 Pu-240 0.1000E+01 none 0.0000E+00 none 0.0000E+00 End of Nuclear Inventory File

DO00040-119, Revision 0 Page 48 of 70 RADTRAD Release Fraction Files

D000040-119, Revision 0 Page 49 of 70 Release Fraction and Timing Altered for Locked Rotor 0 - 24 Hrs. DESIGN PWR, NUREG-1465, Tables 3.12 & 3.13, June 1992 Duration (h):

2.0000E+00 6.0000E+00 1.6000E+01 0.0000E+00 Noble Gases:

8.3300E-02 2.5000E-01 6.6700E-01 0.0000E+00 Iodine:

8.3400E-02 2.5000E-01 6.6700E-01 0.0000E+00 Cesium:

8.3800E-02 2.5100E-01 6.6600E-01 0.0000E+00 Tellurium:

0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00 Strontium:

0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00 Barium:

0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00 Ruthenium:

0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00 Cerium:

0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00 Lanthanum:

0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00 Non-Radioactive Aerosols (kg):

0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00 End of Release File

DC00040-119, Revision 0 Page 50 of 70 Conservative Case RADTRAD Input/Output File

D000040-119, Revision 0 Page 51 of 70 Conservative Case - 1.0 GPM (0 - 24 Hours)

RADTRAD Version 3.03 (Spring 2001) run on 11/08/2011 at 11:40:49 File information Plant file

= C:\\Program Files\\radtrad3 03\\VCS TSC\\Locked Rotor\\l GPM-Conservative Case.psf Inventory file

= C:\\Program Files\\radtrad3_03\\VCS TSC\\Locked Rotor\\0-24 Hr Design.nif Release file

= C:\\Program Files\\radtrad3 03\\VCS TSC\\Locked Rotor\\0-24 Hr Design.rft Dose Conversion file = C:\\Program Files\\radtrad3 03\\VCS TSC\\Locked Rotor\\Fgrll&12.inp Radtrad 3.03 4/15/2001 1 GPM Conservative Case Nuclide Inventory File:

C:\\Program Files\\radtrad3-03\\VCS TSC\\Locked Rotor\\0-24 Hr Design.nif Plant Power Level:

1.0453E+00 Compartments:

3 Compartment 1:

Source Volume 3

1.0000E+04 0

0 0

Compartment 2:

Environment 2

0.0000E+00 0

D000040-119, Revision 0 Page 52 of 70 1

2.0700E+05 0

0 0

Pathways:

4 Pathway 1:

Source Volume to Environment 1

2 2

Pathway 2:

Environment to TSC - Makeup 2

3 2

Pathway 3:

Environment to TSC - Inleakage 2

3 2

Pathway 4:

TSC to Environment - Exhaust 3

2 2

End of Plant Model File Scenario Description Name:

Plant Model Filename:

Source Term:

1 1

1.0000E+00 C:\\Program Files\\radtrad3 03\\VCS TSC\\Locked Rotor\\Fgrll&12.inp C:\\Program Files\\radtrad3 03\\VCS TSC\\Locked Rotor\\0-24 Hr Design.rft 0.0000E+00 1

0.0000E+00 9.7000E-01 3.0000E-02 1.0000E+00 Overlying Pool:

0 0.0000E+00 0

0 0

0

D000040-119, Revision 0 Page 53 of 70 Compartments:

3 Compartment 1:

1 1

0 0

0 Compartment 2:

0 1

0 0

0 Compartment 3:

0 1

0 0

0 Pathways:

4 Pathway 1:

0 0

.0 1

2 0.0000E+00 1.0000E+10 0.0000E+00 0.0000E+00 0.0000E+00 7.2000E+02 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00 0

0 0

0 Pathway 2:

0 0

D000040-119, Revision 0 Page 54 of 70 0

1 2

0.0000E+00 1.2000E+04 0.0000E+00 0.0000E+00 0.0000E+00 7.2000E+02 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00 0

0 0

0 Pathway 3:

0 0

0 1

2 0.0000E+00 1.0000E+03 0.0000E+00 0.0000E+00 0.0000E+00 7.2000E+02 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00 0

0 0

0 Pathway 4:

0 0

0 1

2 0.0000E+00 1.3000E+04 0.0000E+00 0.0000E+00 0.0000E+00 7.2000E+02 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00 0

0 0

0 Dose Locations:

3 Location 1:

EAB 2

1 2

0.0000E+00 1.2400E-04 7.2000E+02 0.0000E+00 1

2 0.0000E+00 3.5000E-04

D000040-119, Revision 0 Page 55 of 70 7.2000E+02 0.0000E+00 0

Location 2:

LPZ 2

1 5

0.0000E+00 2.4200E-05 8.0000E+00 1.6800E-05 2.4000E+01 7.5500E-06 9.6000E+01 2.4000E-06 7.2000E+02 0.0000E+00 1

4 0.0000E+00 3.5000E-04 8.0000E+00 1.8000E-04 2.4000E+01 2.3000E-04 7.2000E+02 0.0000E+00 0

Location 3:

TSC 3

0 1

2 0.0000E+00 3.5000E-04 7.2000E+02 0.0000E+00 1

4 0.0000E+00 1.0000E+00 2.4000E+01 6.0000E-01 9.6000E+01 4.0000E-01 7.2000E+02 0.0000E+00 Effective Volume Location:

1 6

0.0000E+00 3.9000E-05 2.0000E+00 3.3000E-05 8.0000E+00 1.6000E-05 2.4000E+01 1.2000E-05 9.6000E+01 8.7000E-06 7.2000E+02 0.0000E+00 Simulation Parameters:

1 0.0000E+00 0.0000E+00 Output Filename:

C:\\Program Files\\radtrad3 03\\VCS TSC\\Locked Rotor\\l GPM-Conservative Case.oO 1

1 1

0 1

End of Scenario File

D000040-119, Revision 0 Page 56 of 70 RADTRAD Version 3.03 (Spring 2001) run on 11/08/2011 at 11:40:49 Plant Description Number of Nuclides = 60 Inventory Power =

1.0000E+00 MWth Plant Power Level =

1.0453E+00 MWth Number of compartments

=

3 Compartment information Compartment number 1 (Source term fraction =

1.0000E+00 Name: Source Volume Compartment volume =

1.0000E+04 (Cubic feet)

Compartment type is Normal Pathways into and out of compartment 1 Exit Pathway Number 1:

Source Volume to Environment Compartment number 2 Name: Environment Compartment type is Environment Pathways into and out of compartment 2 Inlet Pathway Number 1:

Source Volume to Environment Inlet Pathway Number 4:

TSC to Environment - Exhaust Exit Pathway Number 2:

Environment to TSC - Makeup Exit Pathway Number 3:

Environment to TSC - Inleakage Compartment number 3 Name: TSC Compartment volume =

2.0700E+05 (Cubic feet)

Compartment type is Control Room Pathways into and out of compartment 3 Inlet Pathway Number 2:

Environment to TSC - Makeup Inlet Pathway Number 3:

Environment to TSC -

Inleakage Exit Pathway Number 4:

TSC to Environment - Exhaust Total number of pathways =

4

D000040-119, Revision 0 Page 57 of 70 RADTRAD Version 3.03 (Spring 2001) run on 11/08/2011 at 11:40:49 Scenario Description Radioactive Decay is enabled Calculation of Daughters is enabled Release Fractions and Timings GAP EARLY IN-VESSEL LATE RELEASE RELEASE MASS 2.000000 hr 6.0000 hrs 16.0000 hrs (gm)

NOBLES IODINE CESIUM TELLURIUM STRONTIUM BARIUM RUTHENIUM CERIUM LANTHANUM 8.3300E-02 8.3400E-02 8.3800E-02 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00 2.5000E-01 6.6700E-01 2.5000E-01 6.6700E-01 2.5100E-01 6.6600E-01 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00 4.249E+00 2.023E-02 2.804E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 1

Inventory Power =

MWt Nuclide Name Group Specific Inventory (Ci/MWt) half Whole Body life DCF (s)

(Sv-m3/Bq-s)

Inhaled Thyroid (Sv/Bq)

Inhaled Effective (Sv/Bq)

Kr-85 1

1.310E+03 3.383E+08 1.190E-16 0.000E+00 0.000E+00 Kr-85m 1

2.080E+04 1.613E+04 7.480E-15 0.000E+00 0.000E+00 Kr-87 1

3.800E+04 4.578E+03 4.120E-14 0.000E+00 0.000E+00 Kr-88 1

5.140E+04 1.022E+04 1.020E-13 0.000E+00 0.000E+00 Rb-86 3

1.640E+00 1.612E+06 4.810E-15 1.330E-09 1.790E-09 I-131 2

2.000E+03 6.947E+05 1.820E-14 2.920E-07 8.890E-09 I-132 2

1.830E+03 8.280E+03 1.120E-13 1.740E-09 1.030E-10 I-133 2

2.560E+03 7.488E+04 2.940E-14 4.860E-08 1.580E-09 I-134 2

2.740E+03 3.156E+03 1.300E-13 2.880E-10 3.550E-11 I-135 2

2.350E+03 2.380E+04 8.294E-14 8.460E-09 3.320E-10 Xe-133 1

1.320E+05 4.532E+05 1.560E-15 0.000E+00 0.000E+00 Xe-135 1

2.840E+04 3.272E+04 1.190E-14 0.000E+00 0.000E+00 Cs-134 3

3.720E+02 6.507E+07 7.570E-14 1.110E-08 1.250E-08 Cs-136 3

1.150E+02 1.132E+06 1.060E-13 1.730E-09 1.980E-09 Cs-137 3

2.080E+02 9.467E+08 2.725E-14 7.930E-09 8.630E-09 Nuclide Daughter Fraction Daughter Fraction Daughter Fraction Kr-85m Kr-85 0.21 none 0.00 none 0.00 Kr-87 Rb-87 1.00 none 0.00 none 0.00 Kr-88 Rb-88 1.00 none 0.00 none 0.00 1-131 Xe-131m 0.01 none 0.00 none 0.00 I-133 Xe-133m 0.03 Xe-133 0.97 none 0.00 I-135 Xe-135m 0.15 Xe-135 0.85 none 0.00 Xe-135 Cs-135 1.00 none 0.00 none 0.00 Cs-137 Ba-137m 0.95 none 0.00 none 0.00

D000040-119, Revision 0 Page 58 of 70 Iodine fractions Aerosol

=

0.0000E+00 Elemental

=

9.7000E-01 Organic

=

3.0000E-02 COMPARTMENT DATA Compartment number 1:

Source Volume Compartment number 2:

Environment Compartment number 3:

TSC PATHWAY DATA Pathway number 1: Source Volume to Environment Pathway Filter: Removal Data Time (hr)

Flow Rate Filter Efficiencies (%)

(cfm)

Aerosol Elemental Organic 0.0000E+00 1.0000E+10 0.0000E+00 0.0000E+00 0.0000E+00 7.2000E+02 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00 Pathway number 2: Environment to TSC - Makeup Pathway Filter: Removal Data Time (hr)

Flow Rate Filter Efficiencies (%)

(cfm)

Aerosol Elemental Organic 0.0000E+00 1.2000E+04 0.0000E+00 0.0000E+00 0.0000E+00 7.2000E+02 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00 Pathway number 3: Environment to TSC - Inleakage Pathway Filter: Removal Data Time (hr)

Flow Rate Filter Efficiencies (%)

(cfm)

Aerosol Elemental Organic 0.0000E+00 1.0000E+03 0.0000E+00 0.0000E+00 0.0000E+00 7.2000E+02 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00 Pathway number 4: TSC to Environment - Exhaust Pathway Filter: Removal Data Time (hr)

Flow Rate Filter Efficiencies (%)

(cfm)

Aerosol Elemental Organic 0.0000E+00 1.3000E+04 0.0000E+00 0.0000E+00 0.0000E+00 7.2000E+02 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00 LOCATION DATA Location EAB is in compartment 2

D00004O-119, Revision 0 Page 59 of 70 Location X/Q Data Time (hr)

X/Q (s

m^-3) 0.0000E+00 1.2400E-04 7.2000E+02 0.0000E+00 Location Breathing Rate Data Time (hr)

Breathing Rate (m^3

sec^-l) 0.0000E+00 3.5000E-04 7.2000E+02 0.0000E+00 Location LPZ is in compartment 2

Location X/Q Data Time (hr)

X/Q (s

m^-3) 0.0000E+00 2.4200E-05 8.0000E+00 1.6800E-05 2.4000E+01 7.5500E-06 9.6000E+01 2.4000E-06 7.2000E+02 0.0000E+00 Location Breathing Rate Data Time (hr)

Breathing Rate (m^3

sec^-1) 0.0000E+00 8.0000E+00 2.4000E+01 7.2000E+02 3.5000E-04 1.8000E-04 2.3000E-04 0.0000E+00 Location TSC Location X/Q Data is in compartment 3

Time (hr)

X/Q (s

m^-3) 0.0000E+00 3.9000E-05 2.0000E+00 3.3000E-05 8.0000E+00 1.6000E-05 2.4000E+01 1.2000E-05 9.6000E+01 8.7000E-06 7.2000E+02 0.0000E+00 Location Breathing Rate Data Time (hr)

Breathing Rate (m^3

sec^-l) 0.0000E+00 3.5000E-04 7.2000E+02 0.0000E+00 Location Occupancy Factor Data Time (hr)

Occupancy Factor 0.0000E+00 1.0000E+00 2.4000E+01 6.0000E-01 9.6000E+01 4.0000E-01 7.2000E+02 0.0000E+00 USER SPECIFIED TIME STEP DATA - SUPPLEMENTAL TIME STEPS Time Time step 0.0000E+00 0.0000E+00

DC00040-119, Revision 0 Page 60 of 70 RADTRAD Version 3.03 (Spring 2001) run on 11/08/2011 at 11:40:49 Dose, Detailed model and Detailed Inventory Output Detailed model information at time (H) =

2.0000 EAB Doses:

Time (h) =

2.0000 Whole Body Thyroid TEDE Delta dose (rem) 2.6007E-01 1.0212E+01 6.6951E-01 Accumulated dose (rem) 2.6007E-01 1.0212E+01 6.6951E-01 LPZ Doses:

Time (h) =

2.0000 Whole Body Thyroid TEDE Delta dose (rem) 5.0756E-02 1.9929E+00 1.3066E-01 Accumulated dose (rem) 5.0756E-02 1.9929E+00 1.3066E-01 TSC Doses:

Time (h) =

2.0000 Whole Body Thyroid TEDE Delta dose (rem) 3.6543E-03 2.7768E+00 1.1500E-01 Accumulated dose (rem) 3.6543E-03 2.7768E+00 1.1500E-01 Source Volume Compartment Nuclide Inventory:

Time (h) =

2.0000 Ci kg Atoms Decay Kr-85

.9.5058E-07 2.4229E-12 1.7166E+13 2.5323E+08 Kr-85m 1.1076E-05 1.3459E-15 9.5353E+09 3.4581E+09 Kr-87 9.2693E-06 3.2724E-16 2.2652E+09 4.4730E+09 Kr-88 2.2892E-05 1.8256E-15 1.2493E+10 7.8616E+09 Rb-86 1.1934E-09 1.4667E-17 1.0271E+08 3.1843E+05 I-131 1.4426E-06 1.1636E-14 5.3491E+10 3.8568E+08 I-132 7.2763E-07 7.0492E-17 3.2160E+08 2.6600E+08 I-133 1.7399E-06 1.5359E-15 6.9544E+09 4.7930E+08 1-134 4.0946E-07 1.5349E-17 6.8981E+07 2.6636E+08 I-135 1.3842E-06 3.9416E-16 1.7583E+09 4.1028E+08

D000040-119, Revision 0 Page 61 of 70 Xe-133 9.4751E-05 5.0620E-13 2.2920E+12 2.5378E+10 Xe-135 1.7878E-05 7.0006E-15 3.1229E+10 5.1166E+09 Cs-134 2.7153E-07 2.0986E-13 9.4315E+ll 7.2338E+07 Cs-136 8.3577E-08 1.1403E-15 5.0495E+09 2.2314E+07 Cs-137 1.5183E-07 1.7456E-12 7.6730E+12 4.0448E+07 Source Volume Transport Group Inventory:

Time (h) =

2.0000 Atmosphere Sump Noble gases (atoms) 1.9513E+13 0.0000E+00 Elemental I (atoms) 6.0717E+10 0.0000E+00 Organic I (atoms) 1.8778E+09 0.0000E+00 Aerosols (kg) 1.9566E-12 0.0000E+00 Dose Effective (Ci/cc) 1-131 (Thyroid) 6.2754E-15 Dose Effective (Ci/cc) 1-131 (ICRP2 Thyroid) 7.2607E-15 Total I

(Ci) 5.7038E-06 Source Volume to Environment Transport Group Inventory:

Time (h) =

2.0000 Pathway Filtered Transported Noble gases (atoms)

Elemental I (atoms) 0.0000E+00 0.0000E+00 5.9602E+21 1.8957E+19 Organic I (atoms)

Aerosols (kg) 0.0000E+00 0.0000E+00 5.8630E+17 5.9708E-04 Detailed model information at time (H) =

8.0000 EAB Doses:

Time (h) =

8.0000 Whole Body Thyroid TEDE Delta dose (rem) 3.2566E-01 2.9302E+01 1.5059E+00 Accumulated dose (rem) 5.8573E-01 3.9514E+01 2.1754E+00 LPZ Doses:

Time (h) =

8.0000 Whole Body Thyroid TEDE Delta dose (rem) 6.3556E-02 5.7186E+00 2.9390E-01 Accumulated dose (rem) 1.1431E-01 7.7115E+00 4.2456E-01 TSC Doses:

Time (h) =

8.0000 Whole Body Thyroid TEDE Delta dose (rem) 4.6909E-03 7.8635E+00 3.2142E-01 Accumulated dose (rem) 8.3452E-03 1.0640E+01 4.3642E-01 Source Volume Compartment Nuclide Inventory:

Time (h) =

8.0000 Ci kg Atoms Decay Kr-85 9.5098E-07 2.4239E-12 1.7173E+13 1.0133E+09 Kr-85m 4.3791E-06 5.3212E-16 3.7700E+09 9.2272E+09 Kr-87 3.5229E-07 1.2437E-17 8.6089E+07 6.6530E+09 Kr-88 5.2950E-06 4.2228E-16 2.8898E+09 1.7470E+10 Rb-86 1.1805E-09 1.4509E-17 1.0160E+08 1.2663E+06 1-131 1.4107E-06 1.1379E-14 5.2309E+10 1.5253E+09 1-132 1.1920E-07 1.1548E-17 5.2684E+07 5.3466E+08

D000040-119, Revision 0 Page 62 of 70 1-133 1.4234E-06 1.2566E-15 5.6896E+09 1.7386E+09 1-134 3.5611E-09 1.3349E-19 5.9993E+05 3.3469E+08 1-135 7.3725E-07 2.0993E-16 9.3647E+08 1.2307E+09 Xe-133 9.1758E-05 4.9021E-13 2.2196E+12 9.9916E+10 Xe-135 1.1632E-05 4.5551E-15 2.0319E+10 1.6736E+10 Cs-134 2.7103E-07 2.0948E-13 9.4144E+11 2.8897E+08 Cs-136 8.2347E-08 1.1236E-15 4.9752E+09 8.8563E+07 Cs-137 1.5159E-07 1.7428E-12 7.6607E+12 1.6160E+08 Source Volume Transport Group Inventory:

Time (h) =

8.0000 Atmosphere Sump Noble gases (atoms) 1.9420E+13 0.0000E+00 Elemental I (atoms) 5.7218E+10 0.0000E+00 Organic I (atoms) 1.7696E+09 0.0000E+00 Aerosols (kg) 1.9534E-12 0.0000E+00 Dose Effective (Ci/cc) 1-131 (Thyroid) 5.8964E-15 Dose Effective (Ci/cc) 1-131 (ICRP2 Thyroid) 6.5574E-15 Total I (Ci) 3.6941E-06 Source Volume to Environment Transport Group Inventory:

Pathway Time (h) =

8.0000 Filtered Transported Noble gases (atoms) 0.0000E+00 5.9819E+22 Elemental I (atoms) 0.0000E+00 4.0144E+19 Organic I (atoms) 0.0000E+00 1.2416E+18 Aerosols (kg) 0.0000E+00 1.3003E-03 Detailed model information at time (H) =

24.0000 EAB Doses:

Time (h) = 24.0000 Whole Body Thyroid TEDE Delta dose (rem) 2.2096E-01 7.1338E+01 3.1393E+00 Accumulated dose (rem) 8.0669E-01 1.1085E+02 5.3148E+00 LPZ Doses:

Time (h) = 24.0000 Whole Body Thyroid TEDE Delta dose (rem) 2.9937E-02 4.9707E+00 2.3328E-01 Accumulated dose (rem) 1.4425E-01 1.2682E+01 6.5784E-01 TSC Doses:

Time (h) = 24.0000 Whole Body Thyroid TEDE Delta dose (rem) 1.5770E-03 9.3790E+00 3.8520E-01 Accumulated dose (rem) 9.9222E-03 2.0019E+01 8.2161E-01 Source Volume Compartment Nuclide Inventory:

Time (h) = 24.0000 Ci kg Atoms Decay Kr-85 9.5139E-07 2.4249E-12 1.7180E+13 3.0409E+09 Kr-85m 3.6855E-07 4.4784E-17 3.1729E+08 1.2682E+10 Kr-87 5.7488E-11 2.0296E-21 1.4049E+04 6.7391E+09 Kr-88 1.0669E-07 8.5089E-18 5.8229E+07 2.0303E+10

DC0004O-119, Revision 0 Page 63 of 70 Rb-86 1.1459E-09 1.4083E-17 9.8618E+07 3.7390E+06 I-131 1.3326E-06 1.0749E-14 4.9412E+10 4.4485E+09 I-132 9.6021E-10 9.3024E-20 4.2440E+05 5.8695E+08 I-133 8.3559E-07 7.3763E-16 3.3399E+09 4.0911E+09 I-135 1.3777E-07 3.9231E-17 1.7500E+08 1.9926E+09 Xe-133 8.4152E-05 4.4957E-13 2.0356E+12 2.8730E+ll Xe-135 3.6179E-06 1.4167E-15 6.3197E+09 3.1382E+10 Cs-134 2.6952E-07 2.0831E-13 9.3617E+ll 8.6354E+08 Cs-136 7.9097E-08 1.0792E-15 4.7789E+09 2.6014E+08 Cs-137 1.5083E-07 1..7340E-12 7.6222E+12 4.8305E+08 Source Volume Transport Group Inventory:

Time (h) =

24.0000 Atmosphere Sump Noble gases (atoms) 1.9223E+13' 0.0000E+00 Elemental I (atoms) 5.1339E+10 0.0000E+00 Organic I (atoms) 1.5878E+09 0.0000E+00 Aerosols (kg) 1.9434E-12 0.0000E+00 Dose Effective (Ci/cc) 1-131 (Thyroid) 5.2111E-15 Dose Effective (Ci/cc) 1-131 (ICRP2 Thyroid) 5.5353E-15 Total I (Ci) 2.3069E-06 Source Volume to Environment Transport Group Inventory:

Time (h) = '24.0000 Pathway Filtered Transported Noble gases (atoms)

Elemental I (atoms) 0.0000E+00 0.0000E+00 1.2927E+23 2.6382E+20 Organic I (atoms)

Aerosols (kg) 0.0000E+00 0.0000E+00 8.1594E+18 3.1662E-03 Detailed model information at time (H) =

96.0000 EAB Doses:

Time (h) = 96.0000 Whole Body Thyroid TEDE Delta dose (rem) 1.2908E-10 6.9886E-08 3.0262E-09 Accumulated dose (rem) 8.0669E-01 1.1085E+02 5.3148E+00 LPZ Doses:

Time (h) =

96.0000 Whole Body Thyroid TEDE Delta dose (rem) 7.8592E-12 2.7962E-09 1.2378E-10 Accumulated dose (rem) 1.4425E-01 1.2682E+01 6.5784E-01 TSC Doses:

Time (h) =

96.0000 Whole Body Thyroid TEDE Delta dose (rem) 1.6752E-05 1.6996E-01 7.0629E-03 Accumulated dose (rem) 9.9390E-03 2.0189E+01 8.2868E-01 Source Volume Compartment Nuclide Inventory:

Time (h) = 96.0000 Ci kg Atoms Decay Source Volume Transport Group Inventory:

D000040-119, Revision 0 Page 64 of 70 Time (h) =

96.0000 Atmosphere Sump Noble gases (atoms) 0.0000E+00 0.0000E+00 Elemental I (atoms) 0.0000E+00 0.0000E+00 Organic I (atoms) 0.0000E+00 0.0000E+00 Aerosols (kg) 0.0000E+00 0.0000E+00 Dose Effective (Ci/cc) 1-131 (Thyroid) 0.0000E+00 Dose Effective (Ci/cc) 1-131 (ICRP2 Thyroid) 0.0000E+00 Total I

(Ci) 0.0000E+00 Source Volume to Environment Transport Group Inventory:

Time (h) =

96.0000 Pathway Filtered Transported Noble gases (atoms) 0.0000E+00 1.2985E+23 Elemental I (atoms) 0.0000E+00 2.6536E+20 Organic I (atoms) 0.0000E+00 8.2071E+18 Aerosols (kg) 0.0000E+00 3.2245E-03 Detailed model information at time (H) = 720.0000 EAB Doses:

Time (h) = 720.0000 Whole Body Thyroid TEDE Delta dose (rem) 0.0000E+00 0.0000E+00 0.0000E+00 Accumulated dose (rem) 8.0669E-01 1.1085E+02 5.3148E+00 LPZ Doses:

Time (h) = 720.0000 Whole Body Thyroid TEDE Delta dose (rem) 0.0000E+00 0.0000E+00 0.0000E+00 Accumulated dose (rem) 1.4425E-01 1.2682E+01 6.5784E-01 TSC Doses:

Time (h) = 720.0000 Whole Body Thyroid TEDE Delta dose (rem) 8.3805-124 1.1988-119 5.4748-121 Accumulated dose (rem) 9.9390E-03 2.0189E+01 8.2868E-01 Source Volume Compartment Nuclide Inventory:

Time (h) = 720.0000 Ci kg Atoms Source Volume Transport Group Inventory:

Decay Time (h) = 720.0000 Atmosphere Sump Noble gases (atoms) 0.0000E+00 0.0000E+00 Elemental I (atoms) 0.0000E+00 0.0000E+00 Organic I (atoms) 0.0000E+00 0.0000E+00 Aerosols (kg) 0.0000E+00 0.0000E+00 Dose Effective (Ci/cc).I-131 (Thyroid) 0.0000E+00 Dose Effective (Ci/cc) 1-131 (ICRP2 Thyroid) 0.0000E+00 Total I (Ci) 0.0000E+00 Source Volume to Environment Transport Group Inventory:

Pathway

DC00040-119, Revision 0 Page 65 of 70 Time (h) = 720.0000 Filtered Transported Noble gases (atoms) 0.0000E+00 1.2985E+23 Elemental I (atoms) 0.0000E+00 2.6536E+20 Organic I (atoms) 0.0000E+00 8.2071E+18 Aerosols (kg) 0.0000E+00 3.2245E-03 837

D000040-119, Revision 0 Page 66 of 70 1-131 Summary Time (hr)

Source Volume 1-131 (Curies)

Environment 1-131 (Curies)

TSC 1-131 (Curies) 0.000 1.4530E-06 4.8431E-02 1.1576E-05 0.401 1.4509E-06 3.4937E+01 4.3085E-03 0.701 1.4493E-06 6.1034E+01 5.1286E-03 1.001 1.4478E-06 8.7103E+01 5.3891E-03 1.301 1.4462E-06 1.1314E+02 5.4691E-03 1.601 1.4446E-06 1.3916E+02 5.4909E-03 1.901 1.4431E-06 1.6514E+02 5.4939E-03 2.000 1.4426E-06 1.7370E+02 5.4933E-03 2.300 1.4399E-06 1.9963E+02 4.9149E-03 2.600 1.4383E-06 2.2553E+02 4.7250E-03 2.900 1.4368E-06 2.5140E+02 4.6603E-03 3.200 1.4352E-06 2.7724E+02 4.6361E-03 3.500 1.4337E-06 3.0306E+02 4.6249E-03 3.800 1.4321E-06 3.2884E+02 4.6179E-03 4.100 1.4306E-06 3.5460E+02 4.6123E-03 4.400 1.4290E-06 3.8033E+02 4.6071E-03 4.700 1.4275E-06 4.0604E+02 4.6021E-03 5.000 1.4260E-06 4.3171E+02 4.5971E-03 5.300 1.4244E-06 4.5736E+02 4.5922E-03 5.600 1.4229E-06 4.8299E+02 4.5872E-03 5.900 1.4214E-06 5.0858E+02 4.5823E-03 6.200 1.4198E-06 5.3414E+02 4.5773E-03 6.500 1.4183E-06 5.5968E+02 4.5724E-03 6.800 1.4168E-06 5.8519E+02 4.5675E-03 7.100 1.4152E-06 6.1068E+02 4.5626E-03 7.400 1.4137E-06 6.3613E+02 4.5577E-03 7.700 1.4122E-06 6.6156E+02 4.5528E-03 8.000 1.4107E-06 6.8696E+02 4.5478E-03 8.300 1.4099E-06 7.1235E+02 2.9591E-03 8.600 1.4083E-06 7.3771E+02 2.4450E-03 8.900 1.4068E-06 7.6304E+02 2.2776E-03 9.200 1.4053E-06 7.8835E+02 2.2220E-03 9.500 1.4038E-06 8.1362E+02 2.2024E-03 9.800 1.4023E-06 8.3887E+02 2.1945E-03 10.100 1.4008E-06 8.6410E+02 2.1904E-03 10.400 1.3993E-06 8.8929E+02 2.1875E-03 24.000 1.3326E-06 2.0020E+03 2.0829E-03 96.000 0.0000E+00 2.0020E+03 2.4008-121 720.000 0.0000E+00 2.0020E+03 0.0000E+00

D000040-119, Revision 0 Page 67 of 70 Cumulative Dose Summary EAB LPZ TSC Time (hr)

Thyroid (rem)

TEDE (rem)

Thyroid (rem)

TEDE (rem)

Thyroid (rem)

TEDE (rem) 0.000 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00 0.401 2.0686E+00 1.4748E-01 4.0372E-01 2.8782E-02 3.1097E-01 1.2977E-02 0.701 3.6088E+00 2.5276E-01 7.0429E-01 4.9330E-02 7.3059E-01 3.0431E-02 1.001 5.1431E+00 3.5426E-01 1.0037E+00 6.9138E-02 1.1923E+00 4.9583E-02 1.301 6.6717E+00 4.5236E-01 1.3021E+00 8.8283E-02 1.6664E+00 6.9200E-02 1.601 8.1948E+00 5.4739E-01 1.5993E+00 1.0683E-01 2.1433E+00 8.8891E-02 1.901 9.7125E+00 6.3966E-01 1.8955E+00 1.2484E-01 2.6199E+00 1.0854E-01 2.000 1.0212E+01 6.6951E-01 1.9929E+00 1.3066E-01 2.7768E+00 1.1500E-01 2.300 1.1721E+01 7.5845E-01 2.2875E+00 1.4802E-01 3.2229E+00 1.3334E-01 2.600 1.3225E+01 8.4517E-01 2.5811E+00 1.6494E-01 3.6375E+00 1.5037E-01 2.900 1.4725E+01 9.2987E-01 2.8737E+00 1.8147E-01 4.0411E+00 1.6693E-01 3.200 1.6219E+01 1.0127E+00 3.1653E+00 1.9764E-01 4.4402E+00 1.8328E-01 3.500 1.7708E+01 1.0939E+00 3.4560E+00 2.1348E-01 4.8371E+00 1.9953E-01 3.800 1.9193E+01 1.1735E+00 3.7458E+00 2.2902E-01 5.2324E+00 2.1570E-01 4.100 2.0673E+01 1.2517E+00 4.0346E+00 2.4427E-01 5.6263E+00 2.3181E-01 4.400 2.2148E+01 1.3285E+00 4.3225E+00 2.5927E-01 6.0189E+00 2.4786E-01 4.700 2.3619E+01 1.4041E+00 4.6095E+00 2.7403E-01 6.4103E+00 2.6384E-01 5.000 2.5085E+01 1.4786E+00 4.8957E+00 2.8856E-01 6.8005E+00 2.7978E-01 5.300 2.6547E+01 1.5520E+00 5.1810E+00 3.0289E-01 7.1896E+00 2.9566E-01 5.600 2.8005E+01 1.6244E+00 5.4654E+00 3.1703E-01 7.5775E+00 3.1148E-01 5.900 2.9458E+01 1.6960E+00 5.7490E+00 3.3099E-01 7.9642E+00 3.2726E-01 6.200 3.0907E+01 1.7666E+00 6.0318E+00 3.4478E-01 8.3498E+00 3.4299E-01 6.500 3.2351E+01 1.8365E+00 6.3137E+00 3.5841E-01 8.7342E+00 3.5867E-01 6.800 3.3792E+01 1.9056E+00 6.5949E+00 3.7190E-01 9.1176E+00 3.7431E-01 7.100 3.5228E+01 1.9740E+00 6.8752E+00 3.8525E-01 9.4999E+00 3.8990E-01 7.400 3.6661E+01 2.0418E+00 7.1548E+00 3.9847E-01 9.8811E+00 4.0545E-01 7.700 3.8089E+01 2.1089E+00 7.4335E+00 4.1158E-01 1.0261E+01 4.2095E-01 8.000 3.9514E+01 2.1754E+00 7.7115E+00 4.2456E-01 1.0640E+01 4.3642E-01 8.300 4.0935E+01 2.2414E+00 7.8105E+00 4.2972E-01 1.0942E+01 4.4871E-01 8.600 4.2352E+01 2.3068E+00 7.9093E+00 4.3481E-01 1.1163E+01 4.5771E-01 8.900 4.3765E+01 2.3718E+00 8.0078E+00 4.3984E-01 1.1357E+01 4.6565E-01 9.200 4.5175E+01 2.4362E+00 8.1060E+00 4.4482E-01 1.1543E+01 4.7322E-01 9.500 4.6581E+01 2.5003E+00 8.2040E+00 4.4975E-01 1.1726E+01 4.8067E-01 9.800 4.7983E+01 2.5639E+00 8.3017E+00 4.5463E-01 1.1907E+01 4.8807E-01 10.100 4.9382E+01 2.6271E+00 8.3991E+00 4.5946E-01 1.2088E+01 4.9544E-01 10.400 5.0777E+01 2.6899E+00 8.4963E+00 4.6425E-01 1.2268E+01 5.0278E-01 24.000 1.1085E+02 5.3148E+00 1.2682E+01 6.5784E-01 2.0019E+01 8.2161E-01 96.000 1.1085E+02 5.3148E+00 1.2682E+01 6.5784E-01 2.0189E+01 8.2868E-01 720.000 1.1085E+02 5.3148E+00 1.2682E+01 6.5784E-01 2.0189E+01 8.2868E-01 Worst Two-Hour Doses EAB Time Whole Body Thyroid TEDE (hr)

(rem)

(rem) 0.0 2.6007E-01 1.0212E+01 6.6951E-01

DO00040-119, Revision 0 Page 68 of 70 Westinghouse Transmittal

CGE-93-0007SGUL (PS-CGE-0807) dated. 3/1/93

DC0004O-119, Revision 0 Page 69 of 70 wesrlaghouse Ehergr system Electric Gowataflon Mr. R. 8. Clary Manager, SC Project South Carolina electric & Gas Co.

P.O. Box Be Jenkinsville, SC 29065 CGV-93-0007SGUL O 1fal am PCO reI

eranc 0,nm 8m 335

?ilt:RrgrPpmslHmi^ 15M MS March 1, 1993 PS-CGE-0807 Attention:

John S. Frick

Subject:

Purchase Order No. Q594687 VCSNS RSG/Uprating: Steam Release for Dose Analysis and Write-up for Source Terms

Dear Mr. Clary:

Attached for your information and reference in Gilbert Commonwealth's analyses for the VCSNS RSG/Uprating are.,

1)

The steam releases and feedwater delivery for use in the dose analysis (documented in Westinghouse internal letter 8T-NSASD-TAI-93-60, 2 pages), and The licensing write-up for the source terms used in the dose analysis (no internal Westinghouse letter, 5 pages).

This information is supplied in response to Mr. Paul Bunker's (Gilbert commonwealth) request and was tolecapied to him an March 1, per the agreed-upon schedule. Please contact D. B. Augustine on 412-374-4219 if you have any questions regarding the attached information.

Very truly yours, Buddy B, Joll Y

Y.

Key Accounts Manager.

Carolinas District Power Systems Field.Saba D. B. Augustine/mm muxouPS40:l

D000040-119, Revision 0 Page 70 of 70 WESTINGHOUSE PROPRIETARY CLASS 2 toss of Ottaite Power Steam Release From Three Steam Generators 447900 Ibm (0-2 hours) 8683300 Ibm (2-8 hours)

Feedwater Delivery To Three Steam Generators 375SOO Ibm (0-2 hours) 841800 ibm (2-8 hours)

Stearne;ne Break Failed Fuel 0%

Steam Release From Two Unfaulted 343700 Ibm (0.2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months )

Steam Generators' 733900 Ibm (2.8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months )

Feedwater Delivery To Two Unfautted 445600 Ibm (0-2 hours)

Steam Generators 721500 Ibm (2-8 hours)

Locked Rotor Failed Fuel 1596 Steam Release From Three Steam Generators 447900 Ibm (0-2 hours) 868300 Itxn (2-8 hours)

Feedwater Delivery To Three Steam Generators 375500 Ibm (0-2 hours) 841800 Ibm (2.8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months )

Rod Election Failed Fuel 1056.

TECHNICAL WORK RECORD Originator:

Michael Waselus,,,, i Date:

07/30/2012 System:

N/A Project Title Verification of D000040-119, RO ECR 50786 Page 1

of 3

1.0 Purpose

This TWR is developed to document the verification of D000040-119, R0, "Reactor Coolant Pump Locked Rotor - TSC."

2.0

Reference:

Calculation D000040-119, R0, "Reactor Coolant Pump Locked Rotor - TSC."

3.0 Discussion

The verification of the calculation is to assure that the calculation has been completed IAW ES-412. The verification is performed IAW ES-110. ES-110, Attachment III is included as Attachment 1 to this TWR.

The verification included check of input, mathematical manipulations, computer code input/output and results. The RADTRAD computer code was utilized in the analysis. RADTRAD is a validated and verified code for nuclear use IAW WorleyParsons QA procedures. There are no assumptions in the calculation requiring future confirmation.

4.0 Comments

Technical comments resulted from the review.

1.

None Editorial comments:

1.

Add FSAR Section 15.4.4 to Affected Documents 2.

Add VC Summer Calculation D000040-100, "Reactor Coolant Pump Locked Rotor - AST", to References.

3.

Add EAB and LPZ doses to Summary of Results to compare to results in D000040-100.

4.

Miscellaneous small typographical changes provided to originator incorporated.

5.

Successfully incorporated David McCreary comments.

5.0 Results

The verifier comments have been resolved and the results of the calculation are acceptable for use.

TECHNICAL WORK RECORD Originator:

Michael Waselus Date:

07/30/2012 System:

N/A Project Title Verification of DC00040-119, RO ECR 50786 Page 2 of 3 ES-110 ATTACHMENT III PAGE 1 OF 2 REVISION 2 VERIFICATION RECORD: CALCULATION Calculation #

DC00040-119 Revision 0 The following questions, as a minimum should be answered for calculation verification.

Yes N/A

q

q q

Have inputs, including codes, standards, regulations, requirements, procedures, data and engineering methodology been correctly selected and applied?

Has the calculation been developed in accordance with applicable station procedures (e.g., ES-412).

Is the plant design basis/criteria maintained?

Have assumptions been identified, especially those requiring later confirmation?

Have references been properly identified and complete?

Have the calculation, results, tables and figures been reviewed with regard to numerical accuracy, units and consistency?

Has the calculation been developed/revised in a clear and understandable manner as to not require recourse to the originator?

Is the output reasonable compared to the input?

Do the diagrams or models depicted represent the physical situation correctly and incorporate necessary features for a correct analysis?

Has the calculation cover page been completed in an accurate manner?

Are the sign conventions used in figures and equations consistent?

Is consistent nomenclature used throughout the calculation (e.g., figures, tables)?

Are symbols used on figures and in the text defined?

Are concurrent in-process revisions been addressed and coordinated with this revision?

Has the Calculation Index been updated?

Additional considerations (see attached TWR)?

TECHNICAL WORK RECORD Originator:

Michael Waselus Date:

07/30/2012 System:

N/A Project Title Verification of D000040-119, RO ECR 50786 Page 3

of 3

ES-110 ATTACHMENT III PAGE 2 OF 2 REVISION 2 VERIFICATION RECORD: CALCULATION Calculation #

D000040-119 Revision 0

.61 The following questions, as a minimum should be answered for calculation verification.

CALCULATIONS UTILIZING COMPUTER PROGRAMS:

Has the program been appropriately defined, including the version?

Is the basic methodology used by the program appropriate for the calculation?

Has the appropriate computer program been used?

Has the calculation been performed within the known limits of the program?

Has the computer program been verified and validated in accordance with SAP-1040? Validated by WorleyParsons.

Has the program been defined, controlled, and benchmarked so that the results reported are traceable to a particular version of the program and a particular set of input data?

Have limits for the program been defined, as appropriate?

Comments have been included and resolved.

Is the Validation Data set for the application complete, and provide repeatable results?

M. M. Waselus/

/w allk^

/

07/30/2012 Verifier's Printed Name Verifier's Signature Date

ES-0110 ATTACHMENT XVI PAGE 1 OF,2'q D !

REVISION 2 g^'^>Z REVIEW CONSIDERATIONS: OWNER'S ACCEPTANCE REVIEW ECR/Document Number: D000040-079, D000040-118 Through -123 Project

Title:

Review of New NOB TSC Dose Calculations The following questions should be considered, as a minimum, during the performance of an Owner's Acceptance Review of vendor developed engineering documents.

Yes N/A

q

q M

Is the technical information/design complete, consistent, and correct for the activity under review?

Were inputs, including codes, standards, and regulatory requirements correctly selected and applied?

Are assumptions necessary to perform the design activity adequately described and reasonable? Where necessary, are the assumptions identified for subsequent re-verification when the detailed design activities are completed?

Is the document/package developed in a clear and understandable manner?

Is the plant design basis/criteria maintained?

Are references properly identified and complete?

Were design considerations from EC-01, Attachment I and II adequately add ressed/incorporated?

q Were technical, design, program or procedure requirements adequately addressed/incorporated?

Have applicable construction and operating experiences been considered?

q Were designs developed in accordance with good engineering practices and established ES guidance documents?

q Have impacted documents, databases (EC-02) and equipment changes been identified?

q Is the document/package developed in accordance with applicable station procedures (e.g., SAP-133, ES-453, ES-455)?

q Is the document/package developed in a clear and understandable manner as to not require recourse to the Originator?

D

ES-0110 ATTACHMENT XVI PAGE 2OF,zq REVISION 2 it-ph Yes N/A 11 Does the design meet interfacing organizations operational/maintenance requirements?

Is technical information adequate to perform the task?

Is the acceptance criteria adequate for the activity under review?

Is the post modification testing adequate to confirm the design?

q Has the 10CFR50.59 Review Process been completed, if required?

For work performed in accordance with VC Summer Nuclear Station Procedures, the procedure forms must be signed by the originator and if not qualified must be co-signed by a qualified person. Check the qualifications of the contractor personnel signing the procedure forms.

Yes No

q El 0

Are contractor personnel signing the VCSNS procedure forms qualified under a vendor qualification program or the VCSNS Nuclear Training Manual for those procedures?

If not have the VCSNS forms been co-signed by a person qualified to the applicable procedure?

Technical Reviews

q Are all technical reviews complete and all comments resolved to the satisfaction of the commenter?

TECHNICAL REVIEW: Check all blocks that apply q Principal Piping Engineer q

Principal Engr Analysis Engineer q

Principal I&C Engineer q

Principal Mechanical Engineer q

Principal Civil Engineer q

Principal PSA Engineer q

Principal Nuclear Fuels Engineer q

Principal Digital Engineer q

Principal Electrical Engineer q

Principal EQ Engineer q

Principal Fire Protection Engineer q

Analysis Engineer Dave McCreary q

q q

4& 1 19,y%

-2 017,

[y 90i Dave McCrea

/

Reviewer's Printed Name Reviewer's Sig ature Date D

Detailed Owners Acceptance Review Notes There are three general comments in all of the dose calculations that should be tl'O Z addressed:

1)

The initiation of the HVAC "Emergency Mode" would only be manually. There are statements in the calculations of SI signal initiation. These should be removed to avoid confusion with the Unit 1 Control Room Habitability Envelope and it's automatic Emergency Mode initiation on an SI signal.

MONSE ree reel e remove 2)

The 1000 cfm allotted for TSC "inleakage" isn't applicable in these scenarios since it is very conservatively assumed that none of the air flowing through the TSC is filtered or recirculated. Therefore, 13,000 cfm of unfiltered outside air is flowing into the TSC and out of the TSC. The discussions should be cleaned up to state that the 13,000 cfm assumed conservatively bounds the expected design flow rate value of 12,000 cfm, and that the doses calculated due to assuming no filtration throughout the accident is very conservative since the TSC filtration capability and isolation times are currently unknown.

SPON crm assume c

ree as noted The text'will be revised to'state "The 13 000.

conservatively ;bounds the expected desrgnR^flow`rate value of he doses are 'conservatively calculated assuming no credit for:

out theraccident Prefer not to state that the TSC filtration capability and isolation times are currently unknown : There is a.P&ID showing flters and we could actually t assumed' an operator action such as 30 minutes ifwe had chosen to do so 3)

Suggest not listing procedure revision levels in the verification T VR's. ES-412 and

-110 have been updated.

3 APO RESPO.N regarding Slsignals wi elefe revision levels" lualcomments"'?

reeto all comnien corrections noted be.lpw Main Steam Line Break, D000040-118, Rev. 0:

1.

Follows same format and methodology as in MSLB dose calculation D000040-099 for CR/EAB/LPZ.

2.

Table of Contents needs to be updated for page number corrections.

3.

Section 5.3 should be revised for the general comments above.

Ec.-ci/o XVI.

Pad L ^b'^

Locked Rotor, DC00040-119, Rev. 0:

i9341c-

^,13o^iz, 4.

Follows same format and methodology as in LR dose calculation DC00040-100 for CR/EAB/LPZ.

Sections 2.16 and 5.10 should be revised for the general comments above.

5.

References 4.11 and 4.14 are the same. Per DI-5.10, Ref. 4.11 should be same reference as used in DC00040-118 Ref. 18.

6.

Reference 4.17 is up to Rev 3... Affects DI-5.9 (10gpm value is now 30gpm), but calc uses the correct mass rate so no change in results.

Fuel Handling Accident, DC00040-120, Rev. 0:

7.

Follows same format and methodology as in Fuel Handling dose calculation DC00040-102 for CR/EAB/LPZ.

8.

Sections 2.15 and 5.12 should be revised for the general comments above.

SGTR, DC00040-121, Rev. 0:

9.

Follows same format and methodology as in SGTR dose calculation D000040-098 for CR/EAB/LPZ.

10.

Section 5.3 should be revised for the general comments above.

LOCH; DC00040-122, Rev. 0:

11.

Follows same format and methodology as in LOCA dose calculation D000040-097 for CR/EAB/LPZ.

12.

Section 5.7 should be revised for the general comments above.

CREA, DC00040-123, Rev. 0:

13.

Follows same format and methodology as in CREA dose calculation D000040-101 for CR/EAB/LPZ.

14.

References 4.11 and 4.12 are the same. Per DI-5.10, Ref. 4.11 should be same reference as used in DC00040-118 Ref. 18.

15.

Section 5.10 should be revised for the general comments above.,

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ML14324A323

Text

Description:

SUMMARY

1.0 INTRODUCTION

4.0 REFERENCES

SUMMARY

1.0 INTRODUCTION

4.0 REFERENCES

SUMMARY

Subject:

Dear Mr. Clary:

1.0 Purpose

Reference:

3.0 Discussion

4.0 Comments

5.0 Results

Title:

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