Text

Calculation LM-0643, Rev 1, Re-analysis of Control Rod Drop Accident (CRDA) Using Alternative Source Terms, Attachment 006
	ML053330364
Person / Time
Site:	Limerick
Issue date:	09/21/2005
From:	Reichert P, Rothstein H; Exelon Generation Co, Exelon Nuclear
To:	; Office of Nuclear Reactor Regulation
References
	LM-0643, Rev 1
	Download: ML053330364 (47)
	v • d • e

ii ADDITIONAL ATTACHMENTS TO 10-10-05 Letter: Supplement to Request for LAR Application of AST Attachment 006 AST - LM-0643 Rev 1 CRDA.

I CALCULATION NO. LM-0643 I REV. NO. 001 I PAGE NO.1 ATTACHMEN1 I Design Antlysis Cover Sheet Design Analysis (Major Revision) I Last Page No. ' 18/ Att. F-2 2

Analysis No' L-W0643 Revision: 1

Title:

3 Ri-analysis of Conitrol Rod Drop Accident (CRDA) Using Acrmative Source Tcrns ECtECR No.*' 04-00003 RevisIon:' 0 Limerick Station(s): 7 Generating Component(s):

Station Unit No.:' 1 &2 Discipline:

SEAO Descrip. CD Code/Keyword:' CRDA SafetylOA Class:" Safety Related System Codae. n 912 Structure: ` NIA CONTROLLED DOCUMENT REFERENCES 1" Document No.: From/To Document No.: rem/To DBD LS08B romrrTo Design Analysis LM-0313 Design Analysis LhM0641 From Frorn Caiculation M-78-01 Fromr Design Analysis LM 0645 From Is this Design Analysis Safeguards Information?" Yes 0 No la I yes, see SY-AA-101-106 Does this Design Analysis contain Unverified Yes N 0 Iyes, Y

Assumptions? 7 Ye ° ATUA~u; This Design Analysis SUPERCEDES: " LM-0643. Rev. 0 entirety.

Description of Revision (list affected pages for partials): 19 This revision Incorporates responses to pertinent NRC Request for Additional Information (RAls) with respect to all Exelon Nuclear Station Alternative Source Term Ucense Amendment Applications. The revisions are also to increase the control room ventilation rate to an artificial bounding value of I control room volume change per minute. Finally, additional assumptions from Regulatory Guide 1.1 83 are included to directly indicate conformance with this Regulatory Guide.

Preparer HaroldRotMslein 9/19/2005 Print Name SWgn g'ame Date Method of Detailed Review Alternat Testing 0 Review:21 0 P a T Reviewer. Paul Reichert ( T192005 Print Name Sign Name Date Review Independent review 0 Peer review 3 Notes: 2 Externalli i k L Approver: " tR &(/ .cqkJ ri _____ _____ ____

Print Name Sign N Date Exelon Pnaet ?v//

Rieviewer v1 A Print Na e Date Is a Supplemental Review Required?" Yes [3 No B if yes' lete Atachment 3 Exelon slIorrI;ca &:NoP cJ~'c7 21 eS Print Name Sign Name Date

CALCULATION NO. LM-0643 I REV. NO. 001 PAGE NO. 2 Table of Contents

1. PURPOSE/OBJECTIVE ........ ............................................................... 3

2. METHODOLOGY AND ACCEPTANCE CRITERIA ........................................................................ 4 2.1. General Description.4 2.2. Core Source Term.4 2.3. Fuel Damage Assessment.4 2.4. Radioactivity Transport.4 2.5. Release Pathways .5 2.6. Dose Conversion Factors .5 2.7. Control Room Dose Model .5 2.8. EAB and LPZ Dose Model .5 2.9. Acceptance Criteria .5

3. ASSUMPTIONS ....................................................................... 9

4. DESIGN INPUT ........................................................................ 10 4.1. X/Q Calculations (Meteorology) ....................................................................... 10 4.2. Plant Data ....... ................................................................. 10 4.3. Control Room Data ....................................................................... 10 4.4. Source Terms ....................................................................... 11

5. REFERENCES ....................................................................... 12

6. CALCULATIONS ....... ................................................................ 13 6.1. Source Term Calculation ....................................................................... 13 6.2. Dose Calculations ....................................................................... 13 6.3. SJAE Release Pathway Dose Calculation - Case 2 ........................... ............................................ 15

7.

SUMMARY

AND CONCLUSIONS ......................................................................... 17

8. OWNER'S ACCEPTANCE REVIEW CHECKLIST FOR EXTERNAL DESIGN ANALYSIS ........................... 18 ATTACHMENTS:

A. Release Fraction Assessment Spreadsheet [2 pgs.]

B. RADTRAD Output File [11 pgs.]

C. RADTRAD Source Term "NIF" Input [10 pgs.]

D. RADTRAD Release Fraction "RFT" Input [I pg.]

E. Steam Jet Air Ejector (SJAE) Path Assessment and formulae [2 pgs.]

F. Computer Disclosure Sheets [2 pgs.]

I CALCULATION NO. LM-0643 I REV. NO. 001 I PAGE NO.3

1. PURPOSEIOBJECTIVE The objective of this calculation is to determine the radiological consequences of a Control Rod Drop Accident (CRDA) based on the use of Alternative Source Terms (AST) as defined in Refs. 1 and 3. The design basis CRDA results in the release of radioactivity to the Condenser.

Analyses herein are also performed consistent with the methodology identified in Ref. 13. This is described in UFSAR Section 15.4.9. This is the basis for the previously approved elimination of the reactor trip and MSIV closure on a high-high radiation signal from the Main Steam Line Radiation Monitor (MSLRM). Per Ref. 13, two cases are analyzed. The first case is based on a reactor trip and MSIV closure resulting in an isolated Condenser that is assumed to exhaust at a rate of 1% per day (this is also consistent with Ref. 1, Appendix C). The second case is for no reactor trip or isolation and continued steam flow, so releases would be through the Steam Jet Air Ejectors (SJAE) and the offgas system charcoal delay beds. This path would eliminate iodine releases and greatly delay noble gas releases allowing for decay even with normal off-gas flow rates.

The potential for other forced flow paths from the Turbine/Condenser have also been evaluated per Ref. 1, Appendix C, to determine if consideration is required. For instance, the CRDA could occur during mechanical vacuum pump (MVP) operation, which, if unisolated, would exhaust unprocessed from the Condenser at a significantly larger rate. The MVP is automatically tripped on a high-high radiation signal from the MSLRMs. Therefore, this path would be closed, as discussed in general in Ref. 13.

Releases via steam flow to the gland sealing system are not considered, as clean steam from the Extraction Steam System is utilized for gland sealing. The turbine gland sealing system, which utilizes steam from the gland steam seal evaporator, provides the means of sealing the turbine shaft glands and valve stems.

I CALCULATION NO. LM-0643 I REV. NO. 001 I PAGE NO. 4 l

2. METHODOLOGY AND ACCEPTANCE CRITERIA 2.1. General Description Following a CRDA, radioisotopes postulated to be released will be transported through the Main Steam Lines (MSLs) directly to the Main Steam Condenser. Two scenarios are evaluated, per Ref.

13. For the case of an isolated condenser, the condenser is assumed to leak into the Turbine Building (TB) at a rate of 1% per day, and subsequently be released to the environment through the North Stack, without filtration, and at that same rate. For the case of a condenser not isolated from the reactor, continued off gas flow to the augmented off gas system is assumed. The dispersion that is modeled for this release pathway is defined by the x/0's derived in Ref. 5. The doses from either accident scenario should not exceed the acceptance criteria of the applicable regulatory guidance (Ref. 1, 6).

2.2. Core Source Term For conservatism, the CRDA core source terms are those associated with a DBA power level of 3527 MWth, as per Ref. 9.

2.3. Fuel Damage Assessment The fuel damage from a CRDA is based on failed cladding in a total of 1200 fuel rods in GE12 or GE14 10x10 fuel in an 87.33 equivalent fuel pin array, with a limiting Peaking Factor (PF) of 1.7.

Based on fuel damage assessments in Ref. 14 and as shown below, this bounds the UFSAR Section 15.4.9 analysis for 8x8 array fuel types with 60 fuel pins per bundle, a total of 850 failed rods and a 1.5 PF; it also bounds GE1l1 or GE1 3 9x9 array fuel types with 74 fuel pins per bundle, a total of 1000 failed rods and a 1.5 PF.

Clad Equivalent Clad Damaged Fuel Pins in Damaged Core Fraction Bundle Type Array Bundle Failed Pins Core Fraction PF with PF Various 8x8 60 850 0.018543 1.5 0.027814 GE11&GE13 9x9 74 1000 0.017688 1.5 0.026532 GE12&GE14 10x10 87.33 1200 0.017986 1.7 0.030575 The fuel damage (number of rods with failed cladding and fuel melting in 0.77% of the failed rods) assumptions correspond to those of Ref. 12. Attachment A shows the parameters and breakdown of the fuel damage and subsequent activity released.

2.4. Radioactivity Transport Release fractions and transport fractions are per Regulatory Guide 1.183, Table 3 and Appendix C, as shown in the spreadsheet in Attachment A to this calculation.

I CALCULATION NO. LM-0643 I REV. NO. 001 I PAGENO. 5 2.5. Release Pathways 2.5.1. Turbine/Condenser 1%per day Leakage The Main Condenser is assumed to leak activity into the Turbine Building (TB) at a rate of 1% per day. This activity is then released, unfiltered, to the environment by way of the North Stack, taking no credit for holdup in the TB. The North Stack is the most conservative release point with respect to the Control Room intake, as the normal release pathway via the South Stack is closer to the intake, with lower x/o's.

2.5.2. Steam Jet Air Ejector Discharge When in operation the Steam Jet Air Ejectors (SJAE) discharge to the augmented off-gas system.

This pathway is assessed in Attachment E, through the use of a spreadsheet crediting elimination of Iodine releases and a delay of noble gas releases by the augmented off-gas system charcoal delay beds.

2.6. Dose Conversion Factors The revised Dose Conversion Factors (DCFs) from the U.S. Federal Guidance Report 11 & 12 (Ref. 10, 11) are used for this analysis. The RADTRAD code uses these values directly from its internal database, and when used in spreadsheet analyses they are manually entered.

2.7. Control Room Dose Model For this analysis, as performed using the RADTRAD code, the LGS Unit 1 & 2 Control Room (CR) is modeled as a closed volume of 126,000 ft3. Although the normal maximum flow into the CR is 2100 cfm, a Control Room changeover rate of 1 CR volume per minute is used for conservatism and to allow for unfiltered inleakage/intake. Flow into the CR is therefore assumed to be 126,000 cfm, and to balance the system for analytical purposes, an equal flow of clean air is considered to leave the CR. No credit is taken for any filtration of flows into the CR.

The air that enters the CR originates from a source that is characterized by a dispersion factor (Z/o), calculated using ARCON96 in Ref. 5. The release into the environment from the Turbine Building is postulated to escape through the North Stack, and from the offgas system. The total dose in the Control Room over the 24-hour period is the result of the released activities that enter through the air intake. No CR intake filtration is credited.

2.8. EAB and LPZ Dose Model The Exclusion Area Boundary (EAB) and Low Population Zone (LPZ) V/a's have been determined in Ref. 5, and are located, respectively, 731 m and 2043m from the postulated release points.

2.9. Acceptance Criteria Radiological doses resulting from a design basis CRDA for a control room operator and a person located at EAB or LPZ are to be less than the regulatory dose limits as given in Table 2.1.

I CALCULATION NO. LM-0643 I REV. NO. 001 I PAGE NO. 6 I Table 2.1 Regulatory Dose Limits I Dose Typ e Control Room (rem) I EAB and LPZ (rem) I I TEDE Dose 5a l 6.3l Notes:

a 10 CFR 50.67 (Ref. 6) b Standard Review Plan 15.0.1 (Ref. 3),

Regulatory Guide 1.183 (Ref. 1)

Direct conformance with the relevant sections of the body of Regulatory Guide 1.183 (such as the Acceptance Criteria provided above) and all of the Assumptions in its Appendix C "Assumptions for Evaluating the Radiological Consequences of a BWR Rod Drop Accident" is provided by this analysis, as shown in the Conformance Matrix Table 2.2.

I CALCULATION NO. LM-0643 I REV. NO. 001 PAGE NO. 7 Table 2.2: Conformance with RG 1.183 Appendix C (Control Rod Drop Accident) -

RG' Limerick.

Section RG Position : Analysis- Comments 1 Assumptions acceptable to the NRC staff regarding core inventory are Conforms Analyses based on 100% of provided in Regulatory Position 3 of this guide. For the rod drop the noble gases and 50% of accident, the release from the breached fuel is based on the estimate of the iodines released from the number of fuel rods breached and the assumption that 10% of the melted fuel. Other releases core inventory of the noble gases and iodines is in the fuel gap. The also based on Regulatory release attributed to fuel melting is based on the fraction of the fuel that Position 3.

reaches or exceeds the initiation temperature for fuel melting and on the assumption that 100% of the noble gases and 50% of the iodines contained in that fraction are released to the reactor coolant.

2 If no or minimal fuel damage is postulated for the limiting event, the Conforms Fuel damage is postulated.

released activity should be the maximum coolant activity (typically 4 Therefore, coolant activity is pCi/gm DE 1-131) allowed by the technical specifications.

_ neglected.

3.1 The activity released from the fuel from either the gap or from fuel Conforms All activity released from the pellets is assumed to be instantaneously mixed in the reactor coolant fuel is assumed to mix within the pressure vessel. instantaneously in the reactor coolant.

3.2 Credit should not be assumed for partitioning in the pressure vessel or Conforms No credit for partitioning is for removal by the steam separators. assumed.

3.3 Of the activity released from the reactor coolant within the pressure Conforms Analyses based on 100% of vessel, 100% of the noble gases, 10% of the iodine, and 1% of the the noble gases, 10% of the remaining radionuclides are assumed to reach the turbine and iodines, and 1% of the condensers. remaining nuclides released from reactor coolant reaching the condenser.

3.4 Of the activity that reaches the turbine and condenser, 100% of the Conforms Analyses based on 100% of noble gases, 10% of the iodine, and 1%of the particulate radionuclides the noble gases, 10% of the are available for release to the environment. The turbine and iodines, and 1% of the condensers leak to the atmosphere as a ground- level release at a rate particulate radionuclides of 1% per day for a period of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , at which time the leakage is released from the condenser assumed to terminate. No credit should be assumed for dilution or to the environment, at a

I CALCULATION NO. LM-0643 I REV. NO. 001 I PAGE NO. 8 l Table 2.2: Conformance with RG 1.183 Appendix C (Control Rod Drop Accident):-

RG Limerick -

Section RG Position - Analysis- Comments-holdup within the turbine building. Radioactive decay during holdup in release rate of 1% per day for the turbine and condenser may be assumed. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Decay is assumed in the condenser, but no dilution there or in the turbine building.

3.5 In lieu of the transport assumptions provided in paragraphs 3.2 through Conforms Paragraphs 3.2 through 3.4 3.4 above, a more mechanistic analysis may be used on a case-by-case above are utilized basis. Such analyses account for the quantity of contaminated steam carried from the pressure vessel to the turbine and condensers based on a review of the minimum transport time from the pressure vessel to the first main steam isolation (MSIV) and considers MSIV closure time.

3.6 The iodine species released from the reactor coolant within the pressure Conforms These assumptions are vessel should be assumed to be 95% Csl as an aerosol, 4.85% utilized.

elemental, and 0.15% organic. The release from the turbine and condenser should be assumed to be 97% elemental and 3% organic.

Foot- The activity assumed in the analysis should be based on the activity Conforms Projected fuel damage is the note 1 associated with the projected fuel damage or the maximum technical limiting case.

specification values, whichever maximizes the radiological consequences. In determining the dose equivalent 1-131 (DE 1-131), only the radioiodine associated with normal operations or iodine spikes should be included. Activity from projected fuel damage should not be included.

Foot- If there are forced flow paths from the turbine or condenser, such as Conforms Forced flow paths are note 2 unisolated motor vacuum pumps or unprocessed air ejectors, the considered and the only leakage rate should be assumed to be the flow rate associated with the applicable path is off gas most limiting of these paths. Credit for collection and processing of release, which is considered.

releases, such as by off gas or standby gas treatment, will be considered on a case-by-case basis.

\ CALCULATION NO. LM-0643 I REV.

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NO. 99 CALCULATION NO. LM-0643

3. ASSUMPTIONS

1. Core inventory was based on a DBA power level of 3527 MWth to account for measurement uncertainty for the Rated Thermal Power Level of 3458 MWth.

2. An average power peaking factor of 1.7 per pin was assumed, as per Ref. 12. 10% of the core inventory of noble gases and iodines are released from the fuel gap (Appendix C of Ref. 1). Release fractions of other nuclide groups contained in the fuel gap are detailed in Table 3 of Regulatory Guide 1.183 (Ref. 1).

3. 0.77% of the failed fuel rods will melt during the CRDA, as per Ref. 12. 100% of noble gases and 50% of the iodines contained in the melted fuel fraction are assumed to be released to the reactor coolant (Appendix C of Ref. 1). Fractions of other nuclides released from the melted fuel are used from Table 1 of Regulatory Guide 1.183 (Ref. 1). Though these are described as LOCA values for fuel melt release, they are used to conservatively supplement for missing guidance in regards to the other nuclide groups.

4. The activity released from the fuel from either the gap or from fuel pellets is assumed to be instantaneously mixed with the reactor coolant within the pressure vessel (Ref.

3).

5. 100% of all noble gases, 10% of the iodines, and 1% of remaining nuclides are transported to the Turbine/Condenser (Ref. 1, 3).

6. Of the activity that reaches the Turbine and Condenser, 100% of the noble gases, 10% of the iodine, and 1% of the particulate nuclides are available for release to the environment. (Appendix C of Ref. 1).

7. The MVP is immediately shutdown due to the automatic isolation function of the MSLRM caused by the high radiation levels following a CRDA (Ref. 9).

8. For the isolated condenser case, all leakage from the main steam turbine condenser leaks to the atmosphere from the worst-case X/Q North Stack unfiltered at a rate of 1% per day, for a period of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (Ref. 1, 3).

9. The Control Room was assumed to have no filtration and an occupancy factor of 1 for the 24-hour duration of the CRDA analysis. The CR air intake is conservatively assumed at 1 air change per minute.

10. Charcoal Delay Beds effectively remove all the iodine and particulate from the SJAE activity release (Ref. 9).

11. As the LGS Control Room has no exterior walls or overlying structures that are less than 2 feet thick concrete, this is considered sufficient to eliminate separate consideration of the radiation shine from the external radioactive plume release.

II CALCULATION CACLTO NO. NO LM-0643

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I REV. NO 000 I REV NO. I PAGE AEO NO. 1 0

4. DESIGN INPUT 4.1. XIQ Calculations (Meteorology)

The CR X/a values input to RADTRAD were taken from the ARCON96 results of the LGS Design CaIc. LM-0641, as performed by Washington Group International (WGI) (Ref. 5).

The x/0's were calculated from the worst-case North Stack release point to the Control Room normal fresh air intake.

The CR atmospheric relative concentrations used are as follows (Ref. 5):

X/. =6.88E-03 sec/mi3 (0-2 hours)

X/. =5.17E-03 sec/mi3 (2-8 hours) x/. =2.04E-03 sec/mi3 (8-24 hours)

The EAB and LPZ PAVAN calculated X/o values input to RADTRAD were also taken from the results of the LGS Design Calc. LM-0641 as performed by WGI (Ref. 5). The EAB/LPZ

'Ya's used are as follows (Ref. 5):

EAB x/. =3.18E-04 sec/M 3 (0-2 hours, applied for 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to ensure finding the peak 2-hour period)

LPZ x/a =5.79E-05 sec/M 3 (0-8 hours)

X/. =4.1OE-05 sec/M 3 (8-24 hours) 4.2. Plant Data

DBA Power Level (Ref. 9) 3527 MWth

Radial Peaking Factor (Ref. 12) 1.7

Number of Failed Fuel Rods (bounding case for 10x10 bundle type)(Ref. 12) 1200

Isotopic Release Fractions, as per Reg. Guide 1.183 (Ref. 1) See Attachment A 4.3. Control Room Data

Volume of Control Room, ft3 (Ref. 7) 126,000

Control Room Intake Flow, scfm (unfiltered) 126,000

I CALCULATION NO. LM-0643 I REV. NO. 001 I PAGE NO. 11 4.4. Source Terms The AST values used in this analysis were derived using guidance outlined in Reg. Guide 1.183. A list of 60 core isotopic nuclides and their curie per megawatt activities was extracted from Attachment A of Calculation LM-0645 (Ref. 8) for input into the RADTRAD "NIF" (see Attachment C). The release fractions associated with all of these nuclide groups, as detailed in Regulatory Guide 1.183, were applied to their given groups in Attachment A, and subsequently input into the RADTRAD "RTF", as seen in Attachment D. RADTRAD uses these two files combined with the power of 3527 MWth (Ref. 9) to develop the source terms for this CRDA.

I CALCULATION NO. LM-0643 l REV. NO. 001 I PAGE NO. 12 l

5. REFERENCES

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

2. Design Baseline Document DBD L-S-08B, "Control Room HVAC System", Revision 10.

3. USNRC SRP 15.0.1, Rev. 0, Radiological Consequences Using Alternate Source Terms.

4. NUREG/CR-6604, "RADTRAD: A Simplified Model for RADionuclide Transport and Removal And Dose Estimation", April1998, and Supplements 1, June 1999, and 2, October 2002.

5. LGS Design Analysis LM-0641, "Calculation of Alternative Source Term Onsite and Offsite x/ 0 Values", Rev. 0.

6. 10 CFR 50.67, "Accident Source Term".

7. LGS Calculation No. M-78-01, "Control Room Area - Room Volume", Rev. 6.

8. LGS Design Analysis LM-0645, "Re-analysis of Fuel Handling Accident (FHA) Using Alternative Source Terms", Rev. 0.

9. LGS Design Analysis LM-313, "Impact of Power Rerate on Control Rod Drop Accident Doses And Activities", Rev. 0.

10. U.S. Federal Guidance Report No.11, 'Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion",

1988.

11. U.S. Federal Guidance Report No.12, "External Exposure to Radionuclides in Air, Water, and Soil", 1993.

12. NEDE-31152P, Rev. 7, "General Electric Fuel Bundle Designs", June 2000.

13. NEDO-31400A, "Safety Evaluation for Eliminating the Boiling Water Reactor Main Steam Line Isolation Valve Closure Function and Scram Function of the Main Steam Line Radiation Monitor", October 1992.

14. NEDC-32868P, Rev. 1, "GE14 Compliance With Amendment 22 of NEDE-24011-P-A (GESTAR II)", September 2000.

CALCULATION NO. LM-0643 I REV. NO. 001 PAGE NO. 13l

6. CALCULATIONS 6.1. Source Term Calculation For the RADTRAD calculation, a list of 60 core isotopic nuclides and their activities were extracted from Attachment A of Design Analysis LM-0645 (Ref. 8) for input into the RADTRAD "NIF" (see Attachment C). RADTRAD uses these activities, in curies per megawatt, then applies nuclide release fractions and an input core power to calculate a core source term. The AST release fractions associated with all of these nuclide groups are derived using guidance outlined in Regulatory Guide 1.183, as applied in Attachment A of this calculation. The final gap release and fuel melt release fraction calculated in that attachment, for each nuclide group, is then input into the RADTRAD 'RTF', as seen in Attachment D. RADTRAD applies the input core power of 3527 MWth (Ref. 9) to these two input files to develop the core source term activities for this CRDA.

6.2. Dose Calculations The RADTRAD v. 3.03 computer code is used to determine LGS 1 & 2 CRDA doses at the three dose points cited in Reg. Guide 1.183 (Ref. 1); the Exclusion Area Boundary (EAB), Low Population Zone (LPZ), and Control Room. RADTRAD is a simplified model of RADionuclide Iransport and Removal And Dose Estimation developed for the NRC and endorsed by the NRC as an acceptable methodology for reanalysis of the radiological consequences of design basis accidents.

RADTRAD estimates the releases using the reference AST source terms and assumptions. The RADTRAD code uses a combination of tables and/or numerical models of source term reduction phenomena to determine the time-dependent dose at user-specified locations for a given accident scenario. The code system also provides the inventory, decay chain, and dose conversion factor tables needed for the dose calculation. The technical basis for the RADTRAD code is documented in NUREG/CR-6604 (Ref. 4).

The following is a parameter and descriptions listing of inputs into the RADTRAD model for the calculation of the limiting scenario of an isolated condenser with a condenser leak at 1% / day:

A. Compartments

1. Reactor Coolant - This compartment represents the cooling water within the primary containment vessel.

a. Compartment type - Other - since it is not the environment or control room.

b. Volume - 1 ft3 - This nominal value, used to simplify input, is based on there being a fractional leak rate associated with this compartment.

c. Source term fraction - 1.0 - All of the source term is generated in the reactor coolant.

d. Compartment features - none selected.

2. Condenser - This compartment is the internal volume of the steam condenser.

a. Compartment type - Other - since it is not the environment or control room.

b. Volume - 1 ft3 - This nominal value, used to simplify input, is based on there being a fractional leak rate associated with this compartment.

c. Source term fraction - 0.0

d. Compartment features - none selected.

3. Environment

I CALCULATION NO. LM-0643 I REV. NO. 001 I PAGE NO. 14

a. Compartment type - Environment

4. Control Room

a. Compartment type - Control Room

b. Volume - 126,000 ft3 - Ventilated volume.

c. Source term fraction - 0.0

d. Compartment features - none selected.

B. Transfer Pathways

1. Filtered Flow, Reactor Coolant to Condenser

a. From Compartment 1 - Reactor Coolant

b. To Compartment 2 - Condenser

c. Transfer mechanism - "Filter" selected

d. Filter Efficiency Panel - Flow rate - 10 cfm - With the Reactor Coolant volume set to the nominal value of 1 ft3, this flow rate transfers 99.995% of the activity to the Condenser within 1 minute.

e. Filter Efficiency Panel - Filter efficiency is 0.0%, as no filtration is considered for this accident analysis.

2. Filtered Flow, Condenser to Environment

a. From Compartment 2 - Condenser

b. To Compartment 3 - Environment

c. Transfer mechanism - "Filter" selected -

d. Filter Efficiency Panel - Flow rate - 0.000006944 cfm for 0-24 hrs - This conservatively ignores any holdup in the Condenser. This corresponds to activity leakage from the Condenser at a rate of 1% per day for the duration of the accident.

e. Filter Efficiency Panel - Filter efficiency is 0.0%, as no filtration is considered for this accident analysis.

3. Filtered Flow, Environment to Control Room

a. From Compartment 3 - Environment

b. To Compartment 4 - Control Room

c. Transfer mechanism - Filter" selected -

d. Filter Efficiency Panel - Flow rate - 126,000 cfm -Artificially high CR intake flowrate of one air change per minute, to conservatively allow for any unfiltered inleakage, for the duration of the accident.

e. Filter Efficiency Panel - Filter efficiency is 0.0%, as no filtration is considered for this accident analysis.

4. Filtered Flow, Control Room to Environment

a. From Compartment 4 - Control Room

b. To Compartment 3 - Environment

c. Transfer mechanism - Filter" selected -

d. Filter Efficiency Panel - Flow rate - 126,000 cfm for the duration of the accident.

e. Filter Efficiency Panel - Filter efficiency is entered as 100.0% iodine chemical for all time periods. This is the exit from the control room to the environment; the filtration prevents a double counting of the iodine release. Note that the noble gas release will still be re-circulated between the control room and the outside environment.

C. Dose Locations

I CALCULATION NO. LM-0643 I REV. NO. 001 I PAGE NO. 25

1. Exclusion Area Boundary

a. In Compartment 3 - Environment

b. YQ - 3.18E-04 sec/M 3 for 0-4 hrs - This shows the dispersion to the EAB associated with the North Stack release point for the first 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of the CRDA; EAB dose is only calculated for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months as per regulatory guidance; however, it is applied for 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to ensure finding the peak 2-hour period.

c. Breathing Rate - 3.5E-04 m3/sec; this is the Ref. 1 specified breathing rate. This value is entered from time 0 to the end of the accident.

2. Low Population Zone

a. In Compartment 2 - Environment

b. YQ - 5.79E-05 sec/M 3 for 0-8 hrs; 4.10E-05 sec/M 3 for 8-24 hrs; - This shows the dispersion to the LPZ associated with the North Stack release point for the duration of the CRDA.

c. Breathing Rate - 3.5E-04 m3 /sec for 0-8 hrs; 1.8E-04 m3 /sec for 8-24 hrs - this is the Ref. 1 specified breathing rate assuming a time dependant reduction.

3. Control Room

a. In Compartment 3 - Control Room

b. Y/Q - 6.88E-03 sec/M 3 for 0-2 hrs; 5.17E-03 sec/M 3 for 2-8 hrs; 2.04E-03 sec/M 3 for 8-24 hrs; - This shows the dispersion to the CR associated with the North Stack release point for the duration of the CRDA

c. Breathing Rate - 3.5E-04 m3 /sec for 0-24 hrs - this is the Ref 1 specified breathing rate.

d. Occupancy Factor For 0-24 hours; 0.6 thereafter.

D. Source Term

1. The 'Limerick AST Source Terms.nif' file [Attachment C] reflects the LGS core activities, and is modified to reflect the 'Alternate Source Term" activities provided in Ref. 8.

2. The power level of 3527 MWth, as per Section 4.2 above, accounts for uncertainty.

3. There is no credited delay in the release of activity.

4. The "LGS CRDA-release fractions.rff' file [Attachment D] is designed to reflect gap activity fractions per Regulatory Guide 1.183, Appendix C.

The source terms, which are calculated in Section 6.1 above, are input as a separate RADTRAD "NIF" file. This file is included in Attachment C.

WGI has pre-qualified RADTRAD for application to perform such calculations, as documented in the Computer Disclosure Sheet of Attachment F. The new design basis RADTRAD simulations utilized the design input parameters as provided in Section 4.

6.3. SJAE Release Pathway Dose Calculation - Case 2 The calculation of the dose consequence from the SJAE release pathway was performed using the spreadsheet in Attachment E. The SJAE Release Pathway dose is dependent only upon the noble gas source term, because all iodine and particulate nuclides are effectively eliminated by the

J CALCULATION NO. LM-0643 I REV. NO. 001 I PAGE NO. 16 charcoal delay beds. The initial core activity for each noble gas, calculated at 3527 MWth (Ref. 9),

is then multiplied by the total noble gas release fraction calculated in Attachment A. This activity is then decayed for the period of time that it is delayed in the charcoal beds:

a =a 0 e-At where:

ax= nuclide activity after decay period (Ci) ao= initial nuclide activity (Ci) e= exponential constant A= decay constant (hours"I) t= time of delay (hours)

The two noble gases in the source term, Krypton and Xenon, are characterized by different delay periods. The delay periods utilized are representative values from the LGS UFSAR, verified against the Offgas System Design Baseline Document L-S-30, Rev. 3.

When the decayed activities are found, the dose conversion factors (DCF), dispersion factor ('Io),

and (for the control room dose location) geometry factor (per Equation 1 of Ref. 1) are applied.

The resulting doses from the SJAE release pathway at each respective dose location following a CRDA are developed in Attachment E, and are shown in Table 7.1.

I CALCULATION NO. LM-0643 lTREV. NO. 001 I PAGE NO. 17 l

7.

SUMMARY

AND CONCLUSIONS Table 7.1 provides the results from the RADTRAD code, as well as the prescribed dose acceptance criteria.

Table 7.1. RADTRAD Analysis Results and Comparisons to the Acceplance Criteria EAB LPZ CR Prescribed Dose Limits (TEDE)! 6.3 rem/ 6.3 rem! 5 rem!

Basis Document RG 1.183 RG 1.183 10CFR50.67 RADTRAD Analysis Results (1% of the Condenser free volume 0.0447 0.0312 1.52 leakage per day)

SJAE 0.0226 0.00818 0.0221 For the case analyzed in this calculation assuming condenser isolation, no SGTS, and no CR filtration credited at any point during the 24-hour accident, the limiting CR dose is 1.52 rem TEDE.

This limiting dose is well below the acceptance criteria, so it is verified that no Control Room filtration is needed following a Control Rod Drop Accident.

CALCULATION NO. LM-0643 IREV. NO. 001 IPAGE NO. 18

8. OWNER'S ACCEPTANCE REVIEW CHECKLIST FOR EXTERNAL DESIGN ANALYSIS DESIGN ANALYSIS NO. LM-0643 REV: I Yes No N/A

1. Do assumptions have sufficient rationale?

2. Are assumptions compatible with the way the plant i's El El operated and with the licensing basis? _ X 7) E}/ O El

3. Do the design inputs have sufficient rationale?

4. Are design inputs correct and reasonable?

Are design inputs compatible with the wav-the plart is IT"

[3 El0 El 0

operated and with the licensing basis? I-e rfhr )

6. Are Engineering Judgments clearly documented and justified? El El

'7 Are Engineering Judgments compatible with the way the plant is operated and with the licensing basis? of El

8. Do the results and conclusions satisfy the purpose and

8. objective of the Design Analysis? El El

9. Are the results and conclusions compatible with the w the plant is operated and with the licensing basis? (AV457} El El

10. Does the Design Analysis include the applicable design basis documentation? El El Have any limitations on the use of the results been

11. identified and transmitted to the ap ropriate E1-17 El organizations? A- ApfZUcd J

12. Are there any unverified assumptions? El

13. Do all unverified assumptions have a tracking and closure mechanism in place?

Have all affected design analyses been documented on the

14. Affected Documents List (ADL) for the associated El El Configuration Change?

Do the sources of inputs and analysis methodology used meet current technical requirements and regulatory

15. commitments? (If the input sources or analysis methodology are based on an out-of-date methodology or El El code, additional reconciliation may be required if the site has since committed to a more recent code) 16.

1*

Have vendor supporting technical documents and references (including GE DRFs) been reviewed when necessary? V3/ El El EXELON REVIEWER: , DATE: 9 /,"P,)

// I-Prir)/ Sitnj

A I B I C I D I E I F I G I H 1 I J l K T CRDA AST RADTRAD INPUTS DERIVATION - RELEASE FRACTION THAT REACHES THE CONDENSER USING REGULATORY GUIDE 1.183 APP. C 2 i 3 Constants: I:_

4 Value Description _ Basis I 5 1200 Failed fuel rods - bounding case NEDE 31152P. NEDC-32868P I 6 66720.12 Fuel rods in full core __ NEDE 31152P (87.33-764) I j 7 0.017986 Fraction of rods in core with gap activity release potential 1200/(87.33 pins per assembly x 764 assemblies in core) 8 1.7 Peaking factor I I NEDE 31152P, NEDC-32868P 9 0.030575 Gap activity release potential with peaking 1.7 x .017986 I _

10 0.0077 Fraction of fuel in failed rods assumed to melt NEDE 31152P, NEDC-32868P I 11 0.000235 Melted fuel activity release potential with peaking .030575 x .0077 _ j 12 I 13 =

15 Activity _ I 16 Activity Available for Activity _ _ I 17 Activity Released Release Duration (h): Released Duration (h): jII 18 Released from from 1.OOOOE-03 from 1.OOOOE-03; 1 19 from Gap' Vessel2 Condenser3 Noble Gases: Melted Fuel' Noble Gases:

20 10.00% 100.00% 100.00% 3.0575E-03 90.00% 2.1189E-04 !_!

21 Iodine: Iodine: II 22 10.00% 10.00% 10.00% 3.0575E-05 45.00% 1.0594E-061 !

23 Cesium: j _ijCesium: I j 24 12.00% 1.00% 1.00% 3.6691E-07 _ 20.00% 4.7086E-091 25 Tellurium: _ Tellurium: I 26 0.00% 1.00% 1.00% O.OOOOE+00 _ 5.00% 1.1772E-09i _ _

27 Strontium: Strontium: I I 28 0.00% 1.00% 1.00% O.OOOOE+00 2.00% 4.7086E-10 29 Barium: _ Barium:

30 0.00% 1.00% 1.00% O.OOOOE+00 2.00% 4.7086E-10l 31 Ruthenium: _ Ruthenium: l I 32 0.00% 1.00% 1.00% O.OOOOE+00 i 0.25% 5.8858E-1 1 !

33 Cerium: Cerium: j 34 0.00% 1.00% 1.00% O.OOOOE+00 0.05% 1.1772E-11_ 1 35 Lanthanum: Lanthanum: . I 36 0.00% 1.00% 1.00% O.OOOOE+00 0.02% 4.7086E-121 _

38TIii 39 _ _ I I 40 _ _ I I 41 From Appendix C, paragraph 1. (for Noble gases and Iodine) and Table 3 (for Cesium. an Alkali Metal) of Regulatory Guide 1.183 i _

42 7From Appendix C, paragraph 3.3 of Regulatory Guide 1.183 43 3From Appendix C, paragraph 3.4 of Regulatory Guide 1.183 1 _ _ _ _ _

44 'From Regulatory Guide 1.183, Table 1, Early In-vessel Release Column, with a 100% Noble Gas and 50% Iodine release from fuel l I 45 melting per Appendix C paragraph 1, following subtraction of the gap release fraction. !I li CRDA Release Fractions LNI-0643, Rev. I, Attachment A, Page I of 2

A Il B l C l D T E l F l G lH 1 CRDA AST RADTRAt 2

3 Constants:

4 Value Description Basis 5 1200 Failed fuel rods - bour NEDE 31152P, NEDC-:

6 =87.33-764 Fuel rods in full core NEDE 31152P (87.33-7 7 =A5/A6 Fraction of rods in cor 1200/(87.33 pins per as 8 1_7 _ Peaking factor NEDE 31152P. NEDC-:

9 =A7-A8 Gap activity release F_ 1.7 x.017986 I 10 0.0077 Fraction of fuel in faile NEDE 31152P, NEDC-:

12 =A9'A10 Melted fuel activity rel .030575 x .0077 13 14 15 Activity 1 _6 Activity Available for Activity 17 Activity Released Release Duration (h): Released Duration (h):

18 Released from from 0.001 from 0.001 3 Melted Fuer Noble Gases:

19 from Gap' Vessei' Condenser Noble Gases:

20 0.1 1 1 =AS9-A20-B20-C20 =(1-A20) =A$1 -H20'B20-C20 21 Iodine: iodine:

22 0.1 0.1 0.1 =AS9'A22'B22-C22 0.45 =A$ 1;H22-B22-C22 23 Cesium: , Cesium:

24 0.12 0.01 0.01 =A59'A24-B24'C24 0.2 =AS11 'H24-B24'C24 25 Tellurium: Tellurium:

26 0.01 0.01 =A59-A26'B26-C26 0.05 1=AS11 'H26'B26'C26 27 Strontium: ;Strontium:

28 0.01 0.01 =A$9'A28'B28'C28 0.02 =A51 'H28'B28'C28 29 Barium: Barium:

30 0.01 0.01 =A59-A30-B30-C30 0.02 =A$11-H30-B30tC30 31 Ruthenium: Ruthenium:

32 0.01 0.01 =A$9-A32-B32'C32 0.0025 =A51 1'H32'B32'C32 33 Cerium: Cerium:

340 0.01 0.01 =A$9-A34-B34-C34 0.0005 !=A511-H34-B34'C34 35 Lanthanum: ,Lanthanum:

360 0.01 0.01 =A$9'A36'B36'C36 0.0002 '=A51 -H36-B36'C36 3-8 41 'From Appendix C, p_

42 From Appendix C, pa' 43 'FromAppendix C, p_

44 From Regulatory Guide 1.183. Table 1, Early In-vessel Release Column, with a 100% Noble Gas and 50% lodine release from fuel 45 melting per Appendix I I I I I I I Formnulas LM-0643, Rev. 1, Attachment A, Page 2 of 2

LGS-CRDA.o2 RADTRAD Version 3.03 (Spring 2001) run on 9/19/2005 at 20:29:49 File information Plant file = P:\Users\Nuc\Exelon EOC\Discipline Files\Process\AST\Limerick AST\LGS CRDA\RADTRAD\LGS-CRDA Revl.psf Inventory file = p:\users\nuc\exelon eoc\discipline files\process\ast\limerick ast\limerick ast source terms.nif Release file = p:\users\nuc\exelon eoc\discipline files\process\ast\limerick ast\lgs crda\radtrad\lgs crda-release fractions.rft Dose Conversion file = c:\program files\radtrad3-03\defaults\fgrll&12.inp

1. 1. 1. t# #### ##### # #t #t #####ii #t # #i####t

t #t #t # it ## # #t # # # #

ii#it# #ttii it# #t #t # # # tii# #t #t #

t # #t #t #t # # # # it #

t # it # #t ##i # # it #

t ####i #t #t # ## ###i# #t Radtrad 3.03 4/15/2001 LGS Units 1 & 2 CRDA - No CREF or SGTS Nuclide Inventory File:

p:\users\nuc\exelon eoc\discipline files\process\ast\limerick ast\limerick ast source terms.nif Plant Power Level:

3.5270E+03 Compartments:

4 Compartment 1:

Reactor Coolant 3

1.0000E+00 0

0 0

Compartment 2:

Condenser 3

1.OOOOE+00 0

0 0

Compartment 3:

Environment 2

0.OOOOE+00 0

0 0

0 LM-0643, Rev. 1, Attachment B, Page I of II

LGS-CRDA.o2 0

Compartment 4:

Control Room 1

1.2600E+05 0

0 0

Pathways:

4 Pathway 1:

Reactor Coolant to condenser 1

2 2

Pathway 2:

Condenser to environment 2

3 2

Pathway 3:

Environment to Control Room 3

4 2

Pathway 4:

Control Room to Environment 4

3 2

End of Plant Model File Scenario Description Name:

Plant Model Filename:

Source Term:

1 1 1.0000E+00 c:\program files\radtrad3-03\defaults\fgrll&12.inp p:\users\nuc\exelon eoc\discipline files\process\ast\limerick ast\lgs crda\radtrad\lgs crda-release fractions.rft 0.0000E+00 1

0.OOOOE+00 9.7000E-01 3.OOOOE-02 1.OOOOE+00 Overlying Pool:

0 o.OOOOE+00 0

0 0

0 Compartments:

4 Compartment 1:

0 1

0 0

0 LM-0643, Rev. 1, Attachment B, Page 2 of II

LGS-CRDA.o2 Compartment 2:

0 1

0 0

0 Compartment 3:

0 1

0 0

0 Compartment 4:

0 1

0 0

0 Pathways:

4 Pathway 1:

0 0

0 1

3 O.OOOOE+00 1. OOOOE+01 O.OOOOE+00 O.OOOOE+00 O.OOOOE+00 1.6670E-01 O.OOOOE+00 O.OOOOE+00 O.OOOOE+00 O.OOOOE+00 2.4000E+01 O.OOOOE+00 O.OOOOE+00 O.OOOOE+00 O.OOOOE+00 0

0 0

0 Pathway 2:

0 0

0 1

2 O.OOOOE+00 6.9440E-06 O.OOOOE+00 O.OOOOE+00 O.OOOOE+00 2.4000E+01 O.OOOOE+00 O.OOOOE+00 O.OOOOE+00 O.OOOOE+00 0

0 0

0 LM-0643, Rev. 1, Attachment B, Page 3 of II

LGS-CRDA.o2 Pathway 3:

0 0

0 1

2 O.OOOOE+00 1.2600E+05 O.OOOOE+00 O.OOOOE+00 O.OOOOE+0O 9.6000E+01 O.OOOOE+00 O.OOOOE+00 O.OOOOE+00 O.OOOOE+00 0

0 0

0 Pathway 4:

0 0

0 1

2 O.OOOOE+00 1.2600E+05 1.OOOOE+02 1.OOOOE+02 1.OOOOE+02

9. 6000E+01 O.OOOOE+00 O.OOOOE+00 0.OOOOE+00 O.OOOOE+00 0

0 0

0 Dose Locations:

3 Location 1:

EAB 3

1 2

O.OOOOE+00 3.1800E-04 4.OOOOE+00 O.OOOOE+00 1

2

0. OOOOE+00 3.5000E-04 4.OOOOE+00 O.OOOOE+00 0

Location 2:

LPZ 3

1 3

o .OOOOE+00 5.7900E-05 8.OOOOE+00 4.1000E-05

2. 4000E+01 O.OOOOE+00 1

3 o .OOOOE+00 3.5000E-04 8.0OOOE+00 1.8000E-04 2.4000E+01 O.OOOOE+00 0

Location 3:

Control Room 4

0 1

LM-0643, Rcv. 1, Attachment B, Page 4 of 11

LGS-CRDA.o2 2

O.OOOOE+00 3.5000E-04 9.6000E+01 0.0000E+00 1

3 O.OOOOE+00 1.OOQQE+00 2.4000E+01 6.OOOOE-01 9.6000E+01 O.OOOOE+00 Effective Volume Location:

1 4

O.OQO0E+00 6.8800E-03 2.OOQOE+00 5.1700E-03 8.OOOOE+00 2.0400E-03 2.4000E+01 O.OOOOE+00 Simulation Parameters:

5 O.OOOOE+00 1.OOOQE-04 1.0000E-02 1.OOOOE-03 1.0000E-01 1.OOOOE-02 1.OOQOE+00 1.0000E+00 2.4000E+01 O.OOOOE+00 Output Filename:

P:\Users\Nuc\Exelon EOC\Discipline Files\Process\AST\Limerick AST\LGS CRDA\RADTRAD\LGS-CRDA Revl.oO 1

2 1

0 0

End of Scenario File LM-0643, Rcv. 1, Attachmcnt B, Pagc 5 of I I

LGS-CRDA.o2 RADTRAD Version 3.03 (Spring 2001) run on 9/19/2005 at 20:29:49 Plant Description Number of Nuclides = 60 Inventory Power = 1.OOOOE+00 MWth Plant Power Level = 3.5270E+03 MWth Number of compartments = 4 Compartment information Compartment number 1 (Source term fraction = l.OOOOE+00 Name: Reactor Coolant Compartment volume = 1.OOOOE+00 (Cubic feet)

Compartment type is Normal Pathways into and out of compartment 1 Exit Pathway Number 1: Reactor Coolant to condenser Compartment number 2 Name: Condenser Compartment volume = 1.OOOOE+00 (Cubic feet)

Compartment type is Normal Pathways into and out of compartment 2 Inlet Pathway Number 1: Reactor Coolant to condenser Exit Pathway Number 2: Condenser to environment Compartment number 3 Name: Environment Compartment type is Environment Pathways into and out of compartment 3 Inlet Pathway Number 2: Condenser to environment Inlet Pathway Number 4: Control Room to Environment Exit Pathway Number 3: Environment to Control Room Compartment number 4 Name: Control Room Compartment volume = 1.2600E+05 (Cubic feet)

Compartment type is Control Room Pathways into and out of compartment 4 Inlet Pathway Number 3: Environment to Control Room Exit Pathway Number 4: Control Room to Environment Total number of pathways = 4 LM-0643, Rev. 1, Attachment B, Page 6 of 11

LGS-CRDA.o2

R#i Vrio####03(Spri##g##4201##4ru##n### on91#i#

RADTRAD Version 3.03 (Spring 2001) run on 9, /19/2005 at 20:29:49

1. !i###Oi##i4######?###f##i4######f###4######i####f########## .,i"Xfl #4!.411 411#44 #4!.4141fX.

414!...44 X#11 41

1

# # # ## ## I

1 #f if #14 #41 # #f

f #4 #

41 # ##### # if 41

f #4 #

4# # # #4 if if

! #4 # #! # 41 # if if

if if 4!iif1i##if#if41########if4!#######if!ii4i#!#f1#fi #4!iff#4#ifif4!i#4!if###4!4##4!##4i#if4!

Dose Output

!###Do### # # ######if# ##########################

EAB Doses:

Time (h) = 0.0001 Whole Body Thyroid TEDE Delta dose (rem) 1.8868E-08 2.0183E-07 2.5279E-08 Accumulated dose (rem) 1.8868E-08 2.0183E-07 2.5279E-08 LPZ Doses:

Time (h) = 0.0001 Whole Body Thyroid TEDE Delta dose (rem) 3.4354E-09 3.6748E-08 4.6027E-09 Accumulated dose (rem) 3.4354E-09 3.6748E-08 4.6027E-09 Control Room Doses:

Time (h) = 0.0001 Whole Body Thyroid TEDE Delta dose (rem) 5.5204E-11 1.3080E-08 4 .7067E-10 Accumulated dose (rem) 5.5204E-11 1.3080E-08 4 .7067E-10 EAB Doses:

Time (h) = 0.0002 Whole Body Thyroid TEDE Delta dose (rem) 1.1226E-07 1.1942E-06 1.5020E-07 Accumulated dose (rem) 1.3113E-07 1.3961E-06 1.7547E-07 LPZ Doses:

Time (h) = 0.0002 Whole Body Thyroid TEDE Delta dose (rem) 2.0440E-08 2.1744E-07 2.7347E-08 Accumulated dose (rem) 2.3876E-08 2.5419E-07 3.1950E-08 Control Room Doses:

Time (h) = 0.0002 Whole Body Thyroid TEDE Delta dose (rem) 4.3820E-10 1.0340E-07 3.7226E-09 Accumulated dose (rem) 4.9340E-10 1. 1648E-07 4.1933E-09 EAB Doses:

Time (h) = 0.1667 Whole Body Thyroid TEDE Delta dose (rem) 3.2941E-03 3.4916E-02 4.4026E-03 Accumulated dose (rem) 3.2942E-03 3.4917E-02 4.4028E-03 LPZ Doses:

LM-0643, Rev. 1, Attachment B, Page 7 of II

LGS-CRDA.o2 Time (h) = 0.1667 Whole Body Thyroid TEDE Delta dose (rem) 5.9978E-04 6.3573E-03 8.0161E-04 Accumulated dose (rem) 5.9980E-04 6.3576E-03 8.0165E-04 Control Room Doses:

Time (h) = 0.1667 Whole Body Thyroid TEDE Delta dose (rem) 2.8864E-03 6.7903E-01 2.4444E-02 Accumulated dose (rem) 2.8864E-03 6.7903E-01 2.4444E-02 EAB Doses:

Time (h) = 4.0000 Whole Body Thyroid TEDE Delta dose (rem) 4.7597E-02 7. 8560E-01 7.2358E-02 Accumulated dose (rem) 5.0891E-02 8.2052E-01 7. 6761E-02 LPZ Doses:

Time (h) = 4.0000 Whole Body Thyroid TEDE Delta dose (rem) 8.6663E-03 1.4304E-01 1.3175E-02 Accumulated dose (rem) 9.2661E-03 1.4940E-01 1.3976E-02 Control Room Doses:

Time (h) = 4.0000 Whole Body Thyroid TEDE Delta dose (rem) 4.1952E-02 1.4880E+01 5.1112E-01 Accumulated dose (rem) 4.4838E-02 1.5559E+01 5. 3556E-01 EAB Doses:

Time (h) = 8.0000 Whole Body Thyroid TEDE Delta dose (rem) O.OOOOE+00 O.OOOOE+00 O.OOOOE+00 Accumulatea dose (rem) 5.0891E-02 8.2052E-01 7. 6761E-02 LPZ Doses:

Time (h) = 8.0000 Whole Body Thyroid TEDE Delta dose (rem) 3.8232E-03 1.4112E-01 8.2394E-03 Accumulated dose (rem) 1.3089E-02 2. 9052E-01 2.2216E-02 Control Room Doses:

Time (h) = 8.0000 Whole Body Thyroid TEDE Delta dose (rem) 1.5412E-02 1.2601E+01 4 .0974E-01 Accumulated dose (rem) 6.0250E-02 2.8160E+01 9. 4530E-01 EAB Doses:

Time (h) = 24.0000 Whole Body Thyroid TEDE Delta dose (rem) O.OOOOE+00 O.OOOOE+00 O.OOOOE+00 Accumulated dose (rem) 5.0891E-02 8.2052E-01 7. 6761E-02 LPZ Doses:

Time (h) = 24.0000 Whole Body Thyroid TEDE Delta dose (rem) 3.2995E-03 1. 8436E-01 9.0292E-03 Accumulated dose (rem) 1.6389E-02 4. 7488E-01 3.1245E-02 Control Room Doses:

Time (h) = 24.0000 Whole Body Thyroid TEDE Delta dose (rem) 7.4902E-03 1.7930E+01 5. 6475E-01 Accumulated dose (rem) 6.7740E-02 4.6089E+01 1.5101E+00 LM-0643, Rcv. 1, Attachmcnt B,Page 8 of II

LGS-CRDA.o2 EAB Doses:

Time (h) = 96.0000 Whole Body Thyroid TEDE Delta dose (rem) O.OOOOE+00 O.OOOOE+00 O.OOOOE+00 Accumulated dose (rem) 5.0891E-02 8.2052E-01 7.6761E-02 LPZ Doses:

Time (h) = 96.0000 Whole Body Thyroid TEDE Delta dose (rem) O.OOOOE+00 O.OOOOE+00 O.OOOOE+00 Accumulated dose (rem) 1.6389E-02 4.7488E-01 3.1245E-02 Control Room Doses:

Time (h) = 96.0000 Whole Body Thyroid TEDE Delta dose (rem) 6.9031E-05 3.0817E-01 9. 6115E-03 Accumulated dose (rem) 6.7809E-02 4. 6398E+01 1.5197E+00 I-131 Summary Reactor Coolant Condenser Environment Time (hr) I-131 (Curies) I-131 (Curies) I-131 (Curies) 0.000 2.8124E+03 8.5215E+01 1.1894E-06

0. 000 2.7461E+03 2.5195E+02 8.2270E-06 0.167 1.1283E-40 2.9960E+03 2.0600E-01 0.420 1.1273E-40 2.9930E+03 5.2202E-01 0.670 1.1263E-40 2.9900E+03 8.3361E-01 0.920 1.1253E-40 2.9870E+03 1.1449E+00 1.300 1.1237E-40 2.9824E+03 1.6174E+00

1. 600 1.1225E-40 2.9788E+03 1.9898E+00 1 . 900 1.1213E-40 2.9753E+03 2.3619E+00 2.200 1.1201E-40 2.9717E+03 2.7335E+00 2.500 1.1189E-40 2.9681E+03 3.1046E+00 2.800 1.1177E-40 2.9645E+03 3.4753E+00 3.100 1.1165E-40 2.9610E+03 3.8456E+00 3.400 1.1153E-40 2.9574E+03 4.2154E+00 3.700 1.1141E-40 2.9539E+03 4.5848E+00 4 .000 1.1129E-40 2.9503E+03 4.9537E+00 4 .300 1.1117E-40 2.9468E+03 5.3222E+00 4 . 600 1.1105E-40 2.9432E+03 5.6902E+00 4 .900 1.1093E-40 2.9397E+03 6.0578E+00 5.200 1.1081E-40 2.9362E+03 6.4249E+00 5.500 1.1069E-40 2.9326E+03 6.7917E+00 5.800 1.1057E-40 2.9291E+03 7.1579E+00 6.100 1.1045E-40 2.9256E+03 7.5237E+00 6.400 1.1033E-40 2.9221E+03 7.8891E+00 6.700 1.1021E-40 2.9186E+03 8.2541E+00 7.000 1.1OlOE-40 2.9151E+03 8.6186E+00 7.300 1.0998E-40 2.9115E+03 8.9827E+00 7.600 1.0986E-40 2.9080E+03 9.3463E+00 7.900 1.0974E-40 2.9046E+03 9.7095E+00 8.000 1.0970E-40 2.9034E+03 9.8305E+00 8 . 300 1.0958E-40 2.8999E+03 1.0193E+01

8. 600 1.0946E-40 2.8964E+03 1.0555E+01

8. 900 1.0935E-40 2.8929E+03 1.0917E+01 9.200 1.0923E-40 2.8895E+03 1.1278E+01 9.500 1.0911E-40 2.8860E+03 1.1639E+01 9.800 1.0899E-40 2.8825E+03 1.2000E+01

10. 100 1.0888E-40 2.8790E+03 1.2360E+01 10.400 1.0876E-40 2.8756E+03 1.2719E+01 24 .000 1.0357E-40 2.7230E+03 2.8553E+01

96. 000 7.9969E-41 2.1024E+03 2.8553E+01 LM-0643, Rev. I, Attachment B, Page 9 of II

LGS-CRDA.o2 Control Room Time (hr) I-131 (Curies) 0.000 4.8586E-07 0.000 3.3550E-06 0.167 8.5111E-03 0.420 8.5029E-03 0.670 8.4944E-03 0.920 8.4859E-03 1.300 8.4730E-03

1. 600 8.4628E-03 1.900 8.4526E-03 2.200 6.3441E-03 2.500 6.3365E-03 2.800 6.3289E-03 3.100 6.3213E-03 3.400 6.3137E-03 3.700 6.3061E-03 4.000 6.2985E 4.300 6.2909E-03 4.600 6.2834E-03 4 .900 6.2758E-03 5.200 6.2683E-03 5.500 6.2607E-03 5.800 6.2532E-03 6.100 6.2457E-03 6.400 6.2382E-03 6.700 6.2307E-03 7.000 6.2232E-03 7.300 6.2157E-03

7. 600 6.2082E-03

7. 900 6.2008E-03 8.000 6.1983E-03 8.300 2.4428E-03 8.600 2.4399E-03 8.900 2.4369E-03 9.200 2.434OE-03 9.500 2.4311E-03 9.800 2.4282E-03 10.100 2.4252E-03 10.400 2.4223E-03 24 .000 2.2938E-03

96. 000 O.OOOOE+00 Cumulative Dose Summary EAB LPZ Control Room Time Thyroid TEDE Thyroid TEDE Thyroid TEDE (hr) (rem) (rem) (rem) (rem) (rem) (rem) 0.000 0.OOOOE+00 O.OOOOE+00 O.OOOOE+00 O.OOOOE+00 O.OOOOE+00 O.OOOOE+00 0.000 1.3961E-06 1.7547E-07 2.5419E-07 3. 1950E-08 1.1648E-07 4.1933E-09 0.167 3.4917E-02 4.4028E-03 6.3576E-03 8. 0165E-04 6.7903E-01 2.4444E-02 0.420 8.8329E-02 1.0865E-02 1.6082E-02 1.9783E-03 1.8346E+00 6.5746E-02

0. 670 1.4081E-01 1.6914E-02 2.5638E-02 3.0796E-03 2.9701E+00 1 . 0599E-01 0 . 920 1.9307E-01 2.2662E-02 3.5153E-02 4. 1262E-03 4.1007E+00 1.4575E-01 1.300 2.7212E-01 3.0935E-02 4.9547E-02 5.6325E-03 5.8110E+00 2.0544E-01 1.600 3.3420E-01 3.7074E-02 6.0849E-02 6.7503E-03 7.1540E+00 2.5191E-01 1.900 3.9597E-01 4.2893E-02 7.2096E-02 7.8098E-03 8.4905E+00 2. 9783E-01 2.200 4.5745E-01 4.8422E-02 8.3290E-02 8.8164E-03 9.6558E+00 3.3762E-01 2.500 5.1864E-01 5.3685E-02 9.4432E-02 9.7747E-03 1.0651E+01 3.7138E-01 2.800 5.7956E-01 5.8707E-02 1.0552E-01 1.0689E-02 1.1641E+01 4.0482E-01 3.100 6.4020E-01 6.3508E-02 1. 1656E-01 1. 1563E-02 1.2627E+01 4.3793E-01 3.400 7.0057E-01 6. 8106E-02 1.2756E-01 1.2400E-02 1.3608E+01 4.7075E-01 LM-0643, Rev. I, Attachment B, Page 1O of II

LGS-CRDA.o2 3.700 7. 6068E-01 7.2519E-02 1.3850E-01 1.3204E-02 1.4586E+01 5.0329E-01 4.000 8.2052E-01 7. 6761E-02 1.4940E-01 1.3976E-02 1.5559E+01 5.3556E-01 4.300 8.2052E-01 7. 6761E-02 1. 6025E-01 1.4720E-02 1. 6527E+01 5. 6758E-01 4 .600 8.2052E-01 7. 6761E-02 1.7105E-01 1.5437E-02 1.7492E+01 5. 9935E-01 4.900 8.2052E-01 7. 6761E-02 1.8181E-01 1. 6130E-02 1.8453E+01 6. 3089E-01 5.200 8.2052E-01 7. 6761E-02 1. 9253E-01 1.6800E-02 1.9410E+01 6. 6221E-01 5.500 8.2052E-01 7. 6761E-02 2.0320E-01 1.7449E-02 2.0363E+01 6.9332E-01 5.800 8.2052E-01 7. 6761E-02 2. 1383E-01 1.8079E-02 2.1312E+01 7.2423E-01 6.100 8.2052E-01 7. 6761E-02 2.2441E-01 1.8691E-02 2.2257E+01 7. 5494E-01 6.400 8.2052E-01 7. 6761E-02 2.3496E-01 1.9285E-02 2.3199E+01 7.8546E-01 6.700 8.2052E-01 7. 6761E-02 2.4546E-01 1.9864E-02 2.4137E+01 8.1580E-01 7.000 8.2052E-01 7. 6761E-02 2.5593E-01 2.0429E-02 2.5071E+01 8.4596E-01 7.300 8.2052E-01 7. 6761E-02 2. 6635E-01 2.0979E-02 2.6002E+01 8.7595E-01 7.600 8.2052E-01 7. 6761E-02 2.7673E-01 2. 1517E-02 2.6929E+01 9. 0578E-01 7.900 8.2052E-01 7. 6761E-02 2.8708E-01 2.2043E-02 2.7853E+01 9. 3545E-Ol 8.000 8.2052E-01 7. 6761E-02 2. 9052E-01 2.2216E-02 2.8160E+01 9.4530E-01 8.300 8.2052E-01 7. 6761E-02 2. 9427E-01 2.2467E-02 2.8616E+01 9. 5992E-01 8.600 8.2052E-01 7. 6761E-02 2. 9800E-01 2.2711E-02 2.8977E+01 9.7149E-01 8.900 8.2052E-01 7. 6761E-02 3.0172E-01 2.2949E-02 2.9337E+01 9.8299E-01 9.200 8.2052E-01 7. 6761E-02 3.0543E-01 2.3181E-02 2.9696E+01 9. 9445E-Ol 9.500 8.2052E-01 7. 6761E-02 3.0913E-01 2.3408E-02 3.0053E+01 1.0058E+00 9.800 8.2052E-01 7. 6761E-02 3.1281E-01 2.3629E-02 3.0410E+01 1.0172E+00

10. 100 8.2052E-01 7. 6761E-02 3.1648E-01 2.3845E-02 3.0765E+01 1.0285E+00 10.400 8.2052E-01 7. 6761E-02 3.2013E-01 2.4056E-02 3.1118E+01 1.0397E+00 24.000 8.2052E-01 7.6761E-02 4 .7488E-01 3.1245E-02 4.6089E+01 1.5101E+00

96. 000 8.2052E-01 7.6761E-02 4 .7488E-01 3.1245E-02 4.6398E+01 1.5197E+00 Worst Two-Hour Doses EAB Time Whole Body Thyroid TEDE (hr) (rem) (rem) (rem) 0.0 3.1568E-02 4.1646E-01 4.4736E-02 LM-0643, Rcv. 1, Attachment B, Page 11 of II

Limerick AST Source Terms.nif Nuclide Inventory Name: Source Terms per this calculation Limerick Generating Station (LGS) AST - in Ci/MW Power Level:

0.1000E+01 Nuclides:

60 Nuclide 001:

Co-58 7

0.6117120000E+07 0.5800E+02 0.1529E+03 none O.OOOOE+00 none O.OOOOE+00 none O.OOOOE+00 Nuclide 002:

Co-60 7

0.1663401096E+09 0.6000E+02 0.1830E+03 none O.OOOOE+00 none O.OOOOE+00 none O.OOOOE+00 Nuclide 003:

Kr-85 1

0.3382974720E+09 0.8500E+02 0.3946E+03 none O.OOOOE+00 none O.OOOOE+00 none O.OOOOE+00 Nuclide 004:

Kr-85m 1

0.1612800000E+05 0.8500E+02

0. 8313E+04 Kr-85 0.2100E+00 none O.OOOOE+00 none O.OOOOE+00 Nuclide 005:

Kr-87 1

0.4578000000E+04 0.8700E+02 0.1633E+05 Rb-87 0.1000E+01 none O.OOOOE+00 none O.OOOOE+00 Nuclide 006:

Kr-88 1

0.1022400000E+05 0.8800E+02 0.2303E+05 LM-0643, Rev. 1, Attachment C, Page I of 10

Limerick AST Source Tcrms.nif Rb-88 0.1000E+01 none O.OOOOE+00 none 0.O00OE+OO Nuclide 007:

Rb-86 3

0.1612224000E+07 0.8600E+02 0.6518E+02 none O.OOOOE+00 none O.OOOOE+00 none O.OOOOE+00 Nuclide 008:

Sr-89 5

0.4363200000E+07 0.8900E+02 0.2798E+05 none O.OOOOE+00 none O.OOOOE+00 none O.OOOOE+00 Nuclide 009:

Sr-90 5

0.9189573120E+09 0.9000E+02 0.3178E+04 Y-90 0.1000E+01 none 0.OOOOE+OO none O.OOOOE+00 Nuclide 010:

Sr-91 5

0.34200000000E+05 0.9100E+02 0.3801E+05 Y-91m 0.5800E+00 Y-91 0.4200E+00 none O.OOOOE+00 Nuclide 011:

Sr-92 5

0.9756000000E+04 0.9200E+02 0.4017E+05 Y-92 0.1000E+01 none O.OOOOE+00 none O.OOOOE+00 Nuclide 012:

Y-90 9

0.2304000000E+06 0.9000E+02 0.3272E+04 none O.OOOOE+00 none O.OOOOE+00 none O.OOOOE+00 LM-0643, Rev. I, Attachment C, Page 2 of 10

Limerick AST Source Terms.nif Nuclide 013:

Y-91 9

0.5055264000E+07 0.9100E+02 0.3448E+05 none O.OOOOE+0O none O.OOOOE+00 none O.OOOOE+00 Nuclide 014:

Y-92 9

0.1274400000E+05 0.9200E+02 0.4029E+05 none O.OOOOE+00 none O.OOOOE+00 none O.OOOOE+00 Nuclide 015:

Y-93 9

0.3636000000E+05 0.9300E+02 0.4526E+05 Zr-93 0.1000E+01 none O.OOOOE+00 none O.OOOOE+00 Nuclide 016:

Zr-95 9

0.5527872000E+07 0.9500E+02 0.4489E+05 Nb-95m 0.7000E-02 Nb-95 0.9900E+00 none O.OOOOE+00 Nuclide 017:

Zr-97 9

0.6084000000E+05 0.9700E+02 0.4657E+05 Nb-97m 0.9500E+00 Nb-97 0.5300E-01 none O.OOOOE+00 Nuclide 018:

Nb-95 9

0.3036960000E+07 0.9500E+02 0.4512E+05 none O.OOOOE+00 none O.OOOOE+00 none O.OOOOE+00 Nuclide 019:

Mo-99 7

LM-0643, Rev. I, Attachment C, Page 3 of I0

Limerick AST Source Terms.nif 0.2376000000E+06 O.9900E+02 0.5078+05 Tc-99m 0.8800E+00 Tc-99 0.1200E+00 none 0.OOOOE+00 Nuclide 020:

Tc-99m 7

0.2167200000E+05 0.9900E+02 0.4447E+05 Tc-99 0.1000E+01 none O.OOOOE+00 none O.OOOOE+00 Nuclide 021:

Ru-103 7

0.3393792000E+07 0.1030E+03 0.4202E+05 Rh-103m 0.1000E+01 none O.OOOOE+00 none O.OOOOE+00 Nuclide 022:

Ru-105 7

0.1598400000E+05 0.1050E+03 0.2908E+05 Rh-105 0.1000E+01 none O.OOOOE+00 none O.OOOOE+00 Nuclide 023:

Ru-106 7

0.3181248000E+08 0.1060E+03 0.1730E+05 Rh-106 0.1000E+01 none O.OOOOE+00 none O.OOOOE+00 Nuclide 024:

Rh-105 7

0.1272960000E+06 0.1050E+03 0.2752E+05 none O.OOOOE+00 none O.OOOOE+00 none O.OOOOE+00 Nuclide 025:

Sb-127 4

0.3326400000E+06 0.1270E+03 0.2896E+04 LM-0643, Rev. I, Attachment C, Page 4 of 10

Limerick AST Source Terms.nif Te-127m 0.1800E+00 Te-127 0.8200E+00 none O.OOOOE+00 Nuclide 026:

Sb-129 4

0.1555200000E+05 0.1290E+03 0.8638E+04 Te-129m 0.2200E+00 Te-129 0.7700E+00 none O.OOOOE+00 Nuclide 027:

Te-127 4

0.3366000000E+05 0.1270E+03 0.2873E+04 none 0.OOOOE+00 none 0.OOOOE+00 none O.OOOOE+00 Nuclide 028:

Te-127m 4

0.9417600000E+07 0.1270E+03 0.3855E+03 Te-127 0.9800E+00 none O.OOOOE+00 none O.OOOOE+00 Nuclide 029:

Te-129 4

0.4176000000E+04 0.1290E+03 0.8501E+04 I-129 0.1000E+01 none O.OOOOE+00 none O.OOOOE+OO Nuclide 030:

Te-129m 4

0.2903040000E+07 0.1290E+03 0.1267E+04 Te-129 0.6500E+00 I-129 0.3500E+00 none 0.OOOOE+00 Nuclide 031:

Te-131m 4

0.1080000000E+06 0.1310E+03 0.3869E+04 Te-131 0.2200E+00 I-131 0.7800E+00 none O.OOOOE+00 LM-0643, Rev. 1, Attachment C, Page 5 of 10

Limerick AST Sourcc Tcrms.nif Nuclide 032:

Te-132 4

0.2815200000E+06 0.1320E+03 0.3821E+05 I-132 0.1000E+01 none O.OOOOE+00 none O.OOOOE+00 Nuclide 033:

I-131 2

0.6946560000E+06 0.1310E+03 0.2687E+05 Xe-131m 0.1100E-01 none O.OOOOE+00 none O.OOOOE+00 Nuclide 034:

I-132 2

0.8280000000E+04 0.1320E+03 0.3881E+05 none O.OOOOE+00 none O.OOOOE+00 none 0.OOOOE+00 Nuclide 035:

1-133 2

0.7488000000E+05 0.1330E+03 0.5556E+05 Xe-133m 0.2900E-01 Xe-133 0.9700E+00 none O.OOOOE+00 Nuclide 036:

I-134 2

0.3156000000E+04 0.1340E+03 0.6165E+05 none O.OOOOE+00 none O.OOOOE+00 none O.OOOOE+00 Nuclide 037:

I-135 2

0.2379600000E+05 0.1350E+03 0.5192E+05 Xe-135m 0.1500E+00 Xe-135 0.8500E+00 none O.OOOOE+00 Nuclide 038:

Xe-133 1

LM-0643, Rev. 1, Attachment C, Page 6 of 10

Limerick AST Source Terms.nif 0.4531680000E+06 0.1330E+03 0.5491E+05 none O.OOOOE+00 none O.OOOOE+00 none O.OOOOE+00 Nuclide 039:

Xe-135 1

0.3272400000E+05 0.1350E+03 0.2228E+05 Cs-135 O.lOOOE+01 none O.OOOOE+00 none O.OOOOE+00 Nuclide 040:

Cs-134 3

0.6507177120E+08 0.1340E+03 0.7280E+04 none O.0O00E+00 none O.0O00E+00 none O.OOOOE+00 Nuclide 041:

Cs-136 3

0.1131840000E+07 0.1360E+03 0.2027E+04 none O.OOOOE+00 none O.O000E+00 none O.OOOOE+00 Nuclide 042:

Cs-137 3

0.9467280000E+09 0.1370E+03 0.4538E+04 Ba-137m 0.9500E+00 none O.OOOOE+00 none O.OOOOE+00 Nuclide 043:

Ba-139 6

0.4962000000E+04 0.1390E+03 0.5084E+05 none 0.OOOOE+00 none 0.OOOOE+00 none O.OOOOE+00 Nuclide 044:

Ba-140 6

0.1100736000E+07 0.1400E+03 0.4896E+05 LM-0643, Rev. 1, Attachment C, Page 7 of 10

Limerick AST Source Tcrms.nif La-140 0.1000E+01 none O.OOOOE+00 none O.OOOOE+00 Nuclide 045:

La-140 9

0.1449792000E+06 0.1400E+03 0.5019E+05 none O.OOOOE+00 none O.OOOOE+00 none O.OOOOE+00 Nuclide 046:

La-141 9

0.1414800000E+05 0.1410E+03 0.4640E+05 Ce-141 0.1000E+01 none O.OOOOE+00 none O.OOOOE+00 Nuclide 047:

La-142 9

0.5550000000E+04 0.1420E+03 0.4532E+05 none O.OOOOE+00 none O.OOOOE+00 none O.OOOOE+00 Nuclide 048:

Ce-141 8

0.2808086400E+07 0.1410E+03 0.4492E+05 none O.OOOOE+00 none O.OOOOE+00 none O.OOOOE+00 Nuclide 049:

Ce-143 8

0.1188000000E+06 0.1430E+03 0.4427E+05 Pr-143 0.1000E+01 none O.OOOOE+00 none O.OOOOE+00 Nuclide 050:

Ce-144 8

0.2456352000E+08 0.1440E+03 0.3596E+05 Pr-144m 0.1800E-01 Pr-144 0.9800E+00 none O.OOOOE+00 LM-0643, Rev. 1, Attachment C, Page 8 of 10

Limerick AST Source Tcrms.nif Nuclide 051:

Pr-143 9

0.1171584000E+07 0.1430E+03 0.4293E+05 none O.OOOOE+00 none O.OOOOE+00 none O.OOOOE+00 Nuclide 052:

Nd-147 9

0.9486720000E+06 0.1470E+03 0.1838E+05 Pm-147 0.1000E+01 none O.OOOOE+00 none O.OOOOE+00 Nuclide 053:

Np-239 8

0.2034720000E+06 0.2390E+03 0.5397E+06 Pu-239 O.lOOOE+01 none O.OOOOE+00 none O.OOOOE+00 Nuclide 054:

Pu-238 8

0.2768863824E+10 0.2380E+03 0.1796E+03 U-234 0.1000E+01 none O.OOOOE+00 none O.OOOOE+00 Nuclide 055:

Pu-239 8

0.7594336440E+12 0.2390E+03 0.1200E+02 U-235 0.1000E+01 none O.OOOOE+00 none O.OOOOE+00 Nuclide 056:

Pu-240 8

0.2062920312E+12 0.2400E+03 0.1288E+02 U-236 0.1000E+01 none O.OOOOE+00 none O.OOOOE+00 Nuclide 057:

Pu-241 8

LM-0643, Rev. 1, Attachment C, Page 9 of 10

Limerick AST Source Termns.nif 0.4544294400E+09 0.2410E+03 0.6182E+04 U-237 0.2400E-04 Am-241 0.1000E+01 none O.OOOOE+00 Nuclide 058:

Am-241 9

0.1363919472E+ll 0.241OE+03 0.9528E+01 Np-237 0.1000E+01 none O.OOOOE+00 none O.OOOOE+00 Nuclide 059:

Cm-242 9

0.1406592000E+08 0.2420E+03 0.2388E+04 Pu-238 0.1000E+01 none O.OOOOE+00 none O.OOOOE+00 Nuclide 060:

Cm-244 9

0.5715081360E+09 0.2440E+03 0.2602E+03 Pu-240 0.1000E+01 none O.OOOOE+00 none O.OOOOE+00 End of Nuclear Inventory File LM-0643, Rev. 1, Attachment C, Page 10 of 10

LGS CRDA-rclease fractions.rft Release Fraction and Timing Name:

Limerick Generating Station Duration (h): Control Rod Drop Accident

0. OOlOD+00 0.OOlOD+00 O.OOOOD+00 O.OOOOD+00 Noble Gases:

3.0575E-03 2 .1189E-04 O.OOOOD+00 O.OOOOD+00 Iodine:

3.0575E-05 1.0594E-06 O.OOOOE+00 O.OOOOE+00 Cesium:

3.6691E-07 4.7086E-09 O.OOOOE+00 O.OOOOE+00 Tellurium:

O.OOOOE+00 1.1772E-09 O.OOOOE+00 O.OOOOE+00 Strontium:

O.OOOOE+00 4.7086E-10 O.OOOOE+00 O.OOOOE+00 Barium:

O.OOOOE+00 4.7086E-10 O.OOOOE+00 O.OOOOE+00 Ruthenium:

O.OOOOE+00 5.8858E-1l O.OOOOE+00 O.OOOOE+00 Cerium:

O.OOOOE+00 1.1772E-1l O.OOOOE+00 O.OOOOE+00 Lanthanum:

O.OOOOE+00 4.7086E-12 O.OOOOE+00 O.OOOOE+00 Non-Radioactive Aerosols (kg):

O.OOOOE+00 O.OOOOE+00 O.OOOOE+00 O.OOOOE+00 End of Release File LM-0643, Rcv. 1, Attachment D, Page I of I

A l B C D 0 E F G l H I I J l K 1 Core Initial Core j Decay Release to Release to 2 Half-Live Nominal Source 6 Activity Half-Life Constant Delay Bed Environment 3 Nuclide Isotope Class (seconds) At. Wt. Ci/MW (C (hours) (hours)" (Ci) I (Ci) 4 003: Kr-85 1 _ 3.38E+08 85 3.9460E+02 1.39E+06 9.40E+04 7.38E-06 4.55E+03 i 4.55E+03 5 004: Kr-85m 1 1.61E+04 85 _8.3130E+03 2.93E+07 I4.48E+00 1.55E-01 9.59E+04 3.95E+02 6 005: Kr-87 1 4.58E+03 87 1.6330E+04 5.76E+07 j 1.27E+00 5.45E-01 1.88E+05 7.43E-04 7 006: Kr-88 1 1.02E+04 88 !

2.3030E+04 8.12E+07 2.84E+00 2.44E-01 2.66E+05 4.58E+01 8 038: Xe-133 1 4.53E+05 133 5.4910E+04 1.94E+08 ! 1.26E+02 5.51E-03 6.33E+05 7.08E+03 9 039: Xe-135 i 3.27E+04 135 2.2280E+04 7.86E+07 i 9.09E+00 7.63E-02 2.57E+05 I 2.44E-22 10!i 11I!

12 0.0032694 =CRDA Noble Gas Release Fraction3 I i 13 35.5 =Krypton Holdup in Delay Bed (hrs) 4 _ i 14 816 =Xenon Holdup in Delay Bed (hrs) 4 i 15 3.18E-04 =Vent Stack to EAB XtQ (sec/m3) 5 I !

16 1.15E-04 =Vent Stack to LPZ X/Q (sectm3) 5 ___ I 17 6.88E-03 =Vent Stack to CR X/Q (sectm3) 5 l i 18 4.51E-02 =Control Room Geometry Factor (Murphy Campe Based 2 I_

20 i 21 Release to _ i I 22 Environment DCF 1 EAB Dose LPZ Dose CR Dose _ _

23 Isotope (Ci) (rem TEDE) (rem TEDE) (rem TEDE) _ I 24 Kr-85 4.55E+03 4.403E-04 6.37E-04 2.30E-04 6.22E-04 25 Kr-85m 3.95E+02 2.768E-02 3.47E-03 1.26E-03 3.39E-03 I 26 Kr-87 7.43E-04 1.524E-01 3.60E-08 1.30E-08 3.52E-08 _ _

27 Kr-88 4.58E+01 3.774E-01 5.50E-03 1.99E-03 5.37E-03 28 Xe-133 7.08E+03 5.770E-03 1.30E-02 4.70E-03 1.27E-02 29 Xe-135 2.44E-22 4.400E-02 3.41 E-27 1.23E-27 3.33E-27 _

30II 31 Total Dose (rem TEDE): 2.26E-02 8.18E-03 2.21 E-02 i 32I!

33 I I I i 34 Dose Conversion Factor (rem-m 3 tCurie-second) from Federal Guidance Report 12 per Regulatory Guide 1.183 1 2Reference 1; Equation from K.G. Murphy and K.W. Campe, 13th AEC Air Cleaning Conference, "Nuclear Power Plant 35 Control Room Ventilation System Design for Meeting General Criterion 19", August 1974 !

36 3 Summation from Attachment A j_______l_!

37 LGS UFSAR Eq. 11.3-1 with 75 scfm flow rate, dynamic absorption coefficients of 733 cm3/g for xenon l i 38 and 31.8 cm3/g for krypton, and mass of charcoal adsorber of 321.75 in thousands of pounds from Table 11.3-3 ! I 39 (verified against Offgas System Design Baseline Document, L-S-30, Rev. 3) 1 I 40 Reference 5 worst dispersion factors I I 41 6 Attachment C I I i SJAE Pathway Dose Calculation LM-0643, Rev. 1, Attachment E, Page 1 of 2

A l B [ C [ D [ E i F G l H I J K 1 Core ! Initial Core L ! Decay Release to I Release to 2 Half-Live Nominal Source' j Activity I Half-Life I Constant Delay ed I Environment 3 Nucllde Isotope Class (seconds) At. Wt. Cl/MW I (houn)

._(Cl) I (hours)" (Cl) I (Cl) 4 003: Kr-85 338297472 85 394.6 =F4-3527 '=04/3600 =LN(2yH4 =G4-SAS12 =J4-EXP(-14-SAS13) 5 004: Kr-85m 1 16128 85 8313 =F53527 =D5/3600 i=LN(2VH5 =G5-SA$12 iJ5EXP(-ISSAS13) 6 005: Kr-87 1 4578 87 16330 -=F63527 =613600 I=LN(26 =G6SAS12 =J61EXP(-16 SAS13) 7 006: Kr18 1 0224 88 23030 '=F7-3527 i=D7/3600 I=LN(2yH-17 =G7-SAS12 I=J7-EXP(-17-SAS13) 8 038: Xe-133 1 453168 133 54910 :=F8-3527 =D813600 I=LN(2)H8 =G8-SAS12 '=J8-EXP(-18'SAs14) 9 039: Xe-135 1 32724 135 22280 :=F9-3527 =D913600 !=LN(2Y9 =G9-SSA2 =J9-EXP(-l9-S$S14L 10 _ _ __ _ _ _ __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ I__ I i 11 12 0.0032694 =CRDA Noble Gas Release Fr e I I 13 35.5 =Krypton Holdup in Delay Bed l _

14 816 =Xenon Holdup In Delay Bed (1 _ I 15 0.000318 =Vent Stack to EAB X/O (sec/_ _ _

16 0.000115 =Vent Stack to LPZ X/a (sec/mn i i 17 0.00688 =Vent Stack to CR X/O (sec/m 18 =(126000^0.338y1 173 =Control Room Geometry Fact 19 20 21 Release to I 1

22 Environment DCF EAB Dose LPZ Dose CR Dose I _ _

23 Isotope (Cl) _ (rem TEDE) (rem TEDEL (rem TEDE) I i I 24 Kr-85 =K4 0.0004403 -SB24-$AS15 C24 =SB24-SAS16-SC24 =$B24-SAS17-SC24-SAS18 i i I 25 Kr-85m =K5 0.027676 =SB25-SAS15-C25 =SB25-AS16 SC25 =SB25-SA$17-SC25-SAS18 I 26 Kr487 =K6 0.15244 =SB26-$AS15 C26 =SB26-SAS16-SC26 =SB26-SA$17-SC26-SAS18 27 Kr-88 K7 0.3774 =SB27-SAS15S-C27 *SB27-$A$16SC27 =SB27-SAS17-SC27-SAS18 28 Xe-133 *K8 0.00577 *SB28-SAS15S-C28 =SB28-SAS16-SC28 =SB28-SAS17-SC28-SAS18 29 Xe-135 =K9 0.044 =SB29-$AS15S-C29 =SB29-SAS16-SC29 =SB29-SAS17-SC29-SAS18 30 31 Total Dose (rem TEDE)_ SUM(D24:D29) =SUM(E24:E29) =SUM(F24:F29) 32!

33 34 Dose Conversion Factor (rem-n__iI 2_ Reference 1; Equation from K.G. Murphy and K.W. Campe, 13th AEC Air Cleaning Conference, Nuclear Power Ptant Control Room Ventilation System Design for Meeting General 1

35 ICriterion 19. August 1974 36 3 Summation from Attachment A _ _ j _ I j i iI 37 'lLGS UFSAR Eq 11.3-1 ith 78 _ i i l l i 39 AtReherence 5 worst dispern fz _ _ I_ I l_ _

3 Attachment C IIIII Ii SJAE Pathway Dose Calculation Formulas LM-0643. Rev. 1, Attachment E. Page 2 of 2

Computer Disclosure Sheet Discipline Nuclear Client: Exelon Corporation Date: September 2005 Project: Limerick Generating Station CRDA AST Job No.

Program(s) used Rev No. Rev Date Calculation Set No.: LM-0643, Rev. 1 Attachments A and E spreadsheets N/A N/A Status [ ] Prelim.

[X] Final

[ ] Void WGI Prequalification [ ] Yes (X1 No Run No.

Description:

Analysis

Description:

Spreadsheets used to perform dose assessments for CRDA, as described in calculation.

The attached computer output has been reviewed, the input data checked, And the results approved for release. Input criteria for this analysis were established.

By: On: September 2005 Run by: H. Rothstein Checked by: P. Reichert Do t k Approved by: H. Rothstein J/ i Remarks:

These spreadsheets are applied in a straight-forward manner and were hand checked. Attachments A and E include the spreadsheets in both normal and formula display mode and therefore the spreadsheets are completely documented.

LM-0643, Rev. 1, Attachment F, Page I of 2

Computer Disclosure Sheet Discipline Nuclear Client: Exelon Corporation Date: September 2005 Project: Limerick Generating Station CRDA AST Job No.

Program(s) used: Rev No. Rev Date Calculation Set No.: LM-0643, Rev. 1 RADTRAD 3.03 Runs InAtt. B 0 January 2003 (Prequalification Date)

RADTRAD 3.03 NIF File in Att. C 0 January 2003 Status [ ] Prelim.

RADTRAD 3.03 RFT File in Att. D 0 January 2003 [X] Final

[ ] Void WGI Prequalification [ X ] Yes No Run No.

Description:

Analysis

Description:

RADTRAD output files, where applied to calculations of CRDA dose assessments, as described in calculation.

The attached computer output has been reviewed, the input data checked, And the results approved for release. Input criteria for this analysis were established.

By: On: September 2005 Run by: H. Rothstein .7i (,-

Checked by: P. Reichert /1.

Approved by: H. Rothstein 7/

Remarks:

The RADTRAD computer code is applied in a manner fitting its intended purpose, and well within its operating parameters. All outputs were hand checked. Attachments C & D include the Nuclide Information File and Release Fraction and Timing File used by the RADTRAD code and generated specifically for the Limerick Generating Station. Both were also hand checked for accuracy.

LM-0643, Rev. 1, Attachment F, Page 2 of 2

v t e Letter Sequence Request
TAC:MC2295, Revise Containment Requirements During Handling Irradiated Fuel and Core Alterations (Approved, Closed)
Initiation Request, Request, Request, Request Acceptance Supplement, Supplement, Supplement Administration Meeting, Meeting Questions 8 October 2004 30 June 2005 18 August 2005 3 April 2006 27 April 2006 Answers Results Approval, Approval Other:
MONTHYEAR2005-09-21 [Table View]

ML053330364

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Title:

SUMMARY

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