ML083100205

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Calculation L-003067, Re-analysis of Fuel Handling Accident (FHA) Using Alternative Source Terms, Revision 1, Attachment 9
ML083100205
Person / Time
Site: LaSalle  Constellation icon.png
Issue date: 10/08/2008
From: Rothstein H
Exelon Nuclear
To:
Office of Nuclear Reactor Regulation
References
CC-AA-309-1001, Rev 4 L-003067, Rev. 1
Download: ML083100205 (64)


Text

ATTACHMENT 9 Calculation L-003067, "Re-analysis of Fuel Handling Accident (FHA)

Using Alternative Source Terms," Revision 1

Exek~n CC-AA~-309.i1fli IReision 4 ATTACHMENT I Design Analysis Major Revision Cover Sheet Design Analysis (Major Revision)

Last Page No.-21/Att. F Page F2 Analysis No.:'

L-003067 Revision::

1

Title:

I Re-analysis of Fuel Handling Accident (FHA) Using Alternative Source Terms ECIECR No.:'

EC 352505 Revision: '

0 Station(s):"

LaSaie ouny Component(s):"

Station Unit No.: 1 1&2 NIA Discipline:,

MECH Descrip. Code/Keyword: '*

R01 & R02J F*lA.

DAST, D8A, Dose Safety/QA Class:"

SR System Code: '

ZZ Structure:"

NIA CONTROLLED DOCUMENT REFERENCES "

Document NO.:

romlTo Documwef No.:

FomilTo LA10103 prom NFDE-24011iP-A14-US

[rom L4003128 rom NEDC-32868P rom SEAG 08-00D075 TOOt

-(o DWG. M-17 rom GE Dwg. 10TE1592. Sheet 1

-rom UFSAR. Secwion 15-7 4 From Is this Design Analysis Safeguards Information?

Yes [I No

yes, SY-AA-10t-106 Does this Design Analysis contain Unverified Assumptions? "

Yes [3 No n if yes. ATU/AR This Design Analysis SUPERCEDES: " L-003067 Rev. 0 in its entirety Description of Revision (fist affected pages for partials):

Revised in its entirety. Revision bars are used to show the changes.

Preparer: :H

. Rothstein

/

Method of Review: :

Oetaeild Review 0 Alwattate Caicupations (aacped) 0 resling@

Reiwr:..P.

Reichert i11/

Reiwr onald Gardner

~~-

Review Notes:'

Independent review [

Peer review M All inputS, assumotions. approaches, numerical analyses, and results were independently reviewed and checked. Additionally, a line-by-line teri f the

" aate or was conducted, External Approver:-'

iriAn Tsang

/

Exelon Reviewer:,

J, M.

i.3 P J;L e

independent 3'V Party Review Reqd?

Yes No C]

Exelon Approver:

r Pa, I

/

OrttrotP-Pr4~rý-rmý by Aiwucorr dA>PiId,,v.

Calculation No. L-003067 Exelon Nuclear Rev. No. I Page 1,1 of 33 Revision 7 CC-A Rev(sion 8 I

ATTACHMENT 2 OWNERS ACCEPTANCE REVIEW CHECKLIST FOR EXTERNAL DESIGN ANALYSIS DESIGN ANALYSIS NO. L-003067, Rev. I Page 1.1 1..

Do assumptions have sufficient rationale?

2.

Arc assumptions compatible with the way the plant is operated and with the licensing basis? (

5.)

5j A'T)

3.

Do the design inputs have sufficient rationale?

4.

Are design inputs correct and reasonable?

Are design inputs c9mpatible with the way the plant is operated and with the licensing basis?

0L-kt50,%,

AS%)

6.

Are Engineering Judgments clearly documented and justified?

Are Engineering Judgments compatible with the way the plant is operated and

7.

with the licensing basis?

C.L ul~

kc ArT')4 Do the results and conclusions satisfy the purpose and objective of the Design Analysis?

Are the results and conclusions compatible with the way the plant is operated and with the licensing basis? LtA.st" AST)

10.

Does the Design Analysis include the applicable design basis documentation?

Have any limitations on the use of the results been identified and transmitted to the appropriate organizations? ( pouss ipa+

4e

  • -77 40)
12.

Are there any unverified assumptions?

13.

Do all unverified assumptions have a tracking and closure mechanism in Alace?

1.

Flave all affected design analyses been documented on the Affected Documents List (ADL) for the associated Configuration Change?

Do the sources of inputs and analysis methodology used meet current technical requirements and regulatory conmmitmenls? (If the input sources or

15.

analysis methodology are based on an out-of-date methodology or code, additional reconciliation may be required if the-site has since committed to a more recent code)

Have vendor supporting technical documents and references (including GE DRFs) been reviewed when necessary?

EXELON REVIEWER:

I DATE:

Yes No El El El El El nl 1:1 N/A El El El El El El El M/

El0]

El El El El11 El El El El C1 120,El 2/Ot 081 Io/o 7 Lee I

F

I CALCULATION NO. L-003067 I REV. NO. I

[ PAGE NO. 2 OF 21 1

Table of Contents DESIGN ANALY SIS COV ER SHEET................................................................................................

  • ................. 1 OWNERS ACCEPTANCE REVIEW CHECKLIST FOR EXTERNAL DESIGN ANALYSIS.............. 1.1 TABLE OF CON TEN TS..........................................................................................................................................

2

1.

PURPOSE/OBJECTIVE......................................................................................................................................

3

2.

METHOD OF ANALYSIS AND ACCEPTANCE CRITERIA................................

4 2.1.

Fuel Source Term M odel........................................

4 2.2.

G ap A ctiv ity...............................................................................................................................................

5 2.3.

Pool Decontamination Factor (DF).......................................................................................................

6 2.4.

Release M odel..................................................................

6 2.5.

Control Room M odel..................................................................................................................................

6 2.6.

Dose M odeling...........................................................................................................................................

6 2.6.1.

E A B an d L P Z........ :...................................................................................................................

............. 7 2.6.2.

Control Room........................................................................................................................................

7 2.7.

Acceptance Criteria....................................................................................................................................

7

3.

ASSUM PTIONS.................................................................................................................................................

12

4.

DESIGN INPUT..................................................................................................................................................

14

5.

REFERENCES....................................................................................................................................................

16

6.

CALCULATIONS...............................................................................................................................................

18 6.1.1.

RADTRAD Run Compartment Information..................................................................................

18 6.1.2.

RADTRAD Run Transfer Pathway Information.............................................................................

19 6.1.3.

RADTRAD Run Dose Location Information..................................................................................

19 6.1.4.

RADTRAD Run Source Term & Dose Conversion Factor Information.......................................

20

7.

RESULTS AND CONCLUSIONS.....................................................................................................................

21 Attachments A.

Source Terms Al-A14 B.

RADTRAD Run B1l-B9 C.

FHA RADTRAD Nuclide Information File Cl-Cl0 D.

FHA RADTRAD Release Fraction and Timing File D1-D1 E.

LSCS Fuel Handling Accident Assessment of Limiting Event El-E5 F.

Computer Disclosure Sheets F I-F2

R1 CALCULATION NO. L-003067 REV. NO. 1 PAGE NO.3 OF 21

1.

PURPOSE/OBJECTIVE The purpose of this calculation is to apply Alternative Source Term (AST) methodology to the analysis of the design basis Fuel Handling Accident (FHA) for LaSalle County Station (LSCS)

Units 1 & 2.

Dose consequences are calculated at the Exclusion Area Boundary (EAB), the Low Population Zone (LPZ) and the Control Room. This calculation determines the safety features required to assure that regulatory limits in 10CFR50.67 are met, and is performed in conformance with guidance for analysis of this event provided in Regulatory Guide (RG) 1.183 (Reference [Ref.] 2).

This accident analysis evaluates the movement of fuel that has decayed a minimum of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> since it occupied part of a critical reactor core, to demonstrate that certain available safety R1 features are not required to maintain the accident consequences within acceptance criteria.

The potential for allowing particular doors or penetrations in secondary containment to be left open during movement of fuel with at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> decay will also be identified based on the results of this calculation.

Guidance in Technical Specification Task Force (TSTF) Traveler 51 suggests that a "recently irradiated fuel" parameter be developed to identify when secondary containment integrity features are required for spent fuel movement. Therefore, this calculation demonstrates that 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> decay fuel can be considered as not recently irradiated. In this calculation, recently irradiated fuel would be that which requires one or more of the following safety features:

1. Secondary Containment Integrity to assure that releases are through the plant ventilation stack, which is located on the auxiliary building roof and serves as a single point of release for the reactor building, turbine building, and solid radwaste building ventilation.
2. The Standby Gas Treatment System (SGTS) charcoal adsorber for secondary containment release treatment.
3. The Control Room Area Filtration (CRAF) Makeup subsystem charcoal adsorbers for treated control room pressurization flow as well as the Recirculation Filter subsystem with charcoal adsorbers for airborne radioactivity removal.

All of these systems are expected to be required for operating unit(s) for response to other design basis accidents. The principal benefits of FHA analyses without credit for these systems are:

1. Immediate suspension of movement of irradiated fuel assemblies in the secondary containment would not be required if the fuel being moved (or potentially struck) has sufficient decay of at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and either Technical Specification (TS) Limiting Conditions for Operation (for example) TS 3.6.4.1A; TS 3.6.4.2A; TS 3.6.4.3A; TS 3.7.4.A; or TS 3.7.5.A were to occur.
2.

In the event of a dual unit shutdown, movement of fuel with sufficient decay could be accomplished without operable TS 3.6.4.1, TS 3.6.4.2, TS 3.6.4.3, TS 3.7.4 and TS 3.7.5 systems and with the potential secondary containment opening allowed per Section 7 of this calculation.

Therefore, this calculation supports changes to the current LSCS 1 & 2 Technical Specifications to consider that maintenance of the secondary containment integrity and the operability of emergency filtration systems and subsystems previously required to mitigate the radiological consequences of fuel handling accidents may not be necessary.

CALCULATION NO. L-003067 REV. NO. 1 PAGE NO. 4 OF 21 Based on the discussion in UFSAR (Ref. 1) Section 15.7.4.5 and as per Ref. 13, movement of irradiated fuel will not occur less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the associated reactor fuel has occupied R1 part of a critical reactor core, and therefore, a 24-hour decay period is used as the analyzed condition.

RG 1.183 (Ref. 2) is the basis for these evaluations. Concerning the FHA, this AST guidance has the advantage (compared to previous non-AST analyses) of smaller gap fractions, a larger pool decontamination factor (DF), and dose criteria that replace both the whole body and thyroid dose limits with a limit on Total Effective Dose Equivalent (TEDE).

The other changes from the current UFSAR calculation are listed below.

An offsite dose limit of 6.3 rem TEDE (RG 1.183) is applied instead of the Standard Review Plan (SRP) 15.7.4 value of 25% of the 10CFR100 limits.

A control room dose limit of 5 rem TEDE (10 CFR 50.67(b)(2)(iii)) is applied instead of the 5 rem whole body, or its equivalent (10CFR 50, Appendix A, GDC-1 9).

Design Basis analysis is based on NRC RG 1.183.

The Control Room and offsite doses were recalculated with new limiting XI/Q'S applied to this analysis. With the one exception described in Section 7, Secondary Containment Integrity is assumed, but no Secondary Containment filtration is credited.

CRAF operation is not credited.

Dose Conversion Factors (DCFs) for Immersion and Inhalation are taken from Federal Guidance Reports (FGRs) 12 (Ref. 5) and 11 (Ref. 4), respectively. RG 1.183 cites these DCFs as acceptable current estimates for evaluating the radiological impact of nuclear plant accidents with AST.

This calculation also documents the development of core source terms to be used in this and other design basis accident analyses that involve postulated fuel damage and are being reanalyzed using AST.

2.

METHOD OF ANALYSIS AND ACCEPTANCE CRITERIA Analyses of radiological consequences resulting from a design basis FHA are performed using the guidance for application of AST to this event in RG 1.183, and the approved AST Values for FHA provided in the Ref. 13 "Transmittal of Design Information".

Analyses of radiation transport and dose assessment are performed using RADTRAD v. 3.03.

RADTRAD is a simplified model of RADionuclide Transport 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. The technical basis for the RADTRAD code is documented in NUREG/CR-6604 (Ref. 3). The methodologies significant to this analysis are the dose consequence analysis (NUREG Section 2.3) and the Radioactive Decay Calculations (NUREG Section 2.4). This version of RADTRAD has been pre-qualified for safety related design analysis by Washington Group International per its 10CFR50 Appendix B Quality Assurance program.

2.1.

Fuel Source Term Model

R1I I CALCULATION NO. L-003067 I REV. NO. 1 I PAGE NO. 5 OF 21 As per Ref. 13, the fuel source term is based on the reactor core source terms described in Attachment A. These source terms are bounding for LSCS fuel cycle designs as documented in Attachment A, which repeats the relevant source term information from Ref. 20 for convenience.

R1i The fraction of the core fuel (in the 764 fuel bundle core) damaged is per Ref. 13, Ref. 9, Ref. 10, and Ref. 11 GESTAR II limiting case of damaging 172 fuel pins (based on a "Heavy Mast" design; i.e., the "NF500 mast" in Ref. 10) from GE12 or GE14 1OxI0 fuel bundle arrays with the equivalent of 87.33 pins per bundle, and with all of the damaged fuel assumed to have a limiting Radial Peaking Factor (PF) of 1.7 (per Ref. 13). This analysis is for an assembly and mast drop from a 34 feet maximum height from the refueling platform over the reactor well onto the reactor core, bounding in terms of fuel damage potential. Based on fuel damage assessments in references 11, 14, and 15 as shown in Table 1 below, this bounds all currently used and historical fuel types (including any 7x7 array fuel that may have been used early in the reactor life, now sufficiently decayed that any releases will be bounded by the other fuel types considered below).

GE-Various 8x8 62 124 0.002618 1.5 0.003927 FANP Atrium-9B 9x9 72 131 0.002381 1.5 0.003572 Atrium-10 1Ox10 91 156 0.002244 1.7 0.003815 GE11&GE13 9x9 74 140 0.002476 1.5 0.003714 GE12&GE14*

1Ox10 87.33 172 0.002578 1.7 0.004382 Bounding Assembly type, with Radial Peaking Factor commensurate with full core application.

With a 1.7 radial peaking factor, the associated power of the damaged fuel = 3559 MWth

  • 0.002578
  • 1.7 = 15.597 MWth.

2.2.

Gap Activity This calculation is applicable to fuel whose burnup and power limits are bounded by those specified in RG 1.183, footnote 11. This allows application of the gap activity fractions for Loss of Coolant Accident (LOCA) events per Table 3 of RG 1.183, which are as follows:

5% of the noble gases (excluding Kr-85) 10% of the Kr-85 5% of the iodine inventory (excluding 1-131) 8% of the 1-131 12% of the Alkali metal inventory Because RADTRAD does not allow for application of isotope specific release fractions, the "LaSalle Generating Station AST Source Term.nif' file is modified to accommodate the differential gap activities among the halogen (1-131) and noble gas (Kr-85) gap fractions dictated by RG 1.183 (Ref. 2) shown above. Therefore, the initial activity of isotope 1-131 and Kr-85 are multiplied by 1.6 and 2.0, respectively, in order to accommodate the respective 10% and 8%

release fractions directed by regulatory guidance (Ref. 2).

CALCULATION NO. L-003067 REV. NO. 1 PAGE NO. 6 OF 21 2.3.

Pool Decontamination Factor (DF)

Attachment E provides assessments of worst-case water coverage and fuel damage for FHAs over the reactor well and the spent fuel pool, and demonstrates that the drop over the reactor well is more limiting and therefore the bounding case. This is due to the greater number of fuel rods damaged for the reactor well drop (117 for the spent fuel pool drop vs. 172 for the reactor well drop for the bounding 1 Ox1 0 fuel array, or a ratio of 68.0%), and the fact that the lower than 200 iodine decontamination factor for the reduced water coverage depths applicable (per Ref. 2 and its reference to Ref. 8) to a drop over the spent fuel pool is not significant enough to overcome the fuel damage difference (as per Attachment E, a calculated ratio of 68.8%, greater than the above damage ratio).

2.4.

Release Model Release modeling uses the RADTRAD computer program. The Release Fraction and Timing (RFT) files for this event provide for a rapid release (1.OE-04 hours) of gap activity to the pool water. The gap fraction for noble gas is 5% (except for Kr-85), and for iodine is also 5% (except for 1-131). As discussed above, the normal Nuclide Inventory File (NIF) representing a LSCS core is artificially adjusted to account for the higher than average gap fractions for 1-131 and Kr-85 provided by RG 1.183.

The compartments are the Reactor Building Air Space, the Environment, and the Control Room.

The Reactor Building exhaust rate is set artificially high at 0.1 air changes/minute to assure essentially complete release within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

This accident analysis evaluates the movement of fuel that has decayed a minimum of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> since it occupied part of a critical reactor core, to demonstrate that certain available safety features are not required to maintain consequences within acceptance criteria even for this minimum time as a "recently irradiated fuel" limit. All releases are assumed to be from a worst-case location with Secondary Containment Integrity maintained (with the one exception described in Section 7) but no SGTS filtration is credited.

2.5.

Control Room Model Note: The LSCS Control Room (CR) currently includes both the CR and the Auxiliary Electric Equipment Room (AEER). As stated in Section 7, AEER occupancy is not required in response to an FHA.

The CRAF System is determined to not be required for this event and is not credited. The intake rate is set at an extreme value of 30,000 cfm, which exceeds by about 14% the control room ventilation system purge flow rate of 26,340 cfm (Ref. 21). This is not an expected condition but maximizes the intake rate and the speed at which control room radioactivity concentrations approach outside conditions for conservatism.

2.6.

Dose Modeling Dose models for both onsite and offsite are simplified and meet RG 1.183 requirements. Dose conversion factors are based on Federal Guidance Reports 11 and 12 (Ref. 4, 5). RADTRAD uses the following formulations, integrated numerically over the accident duration:

CALCULATION NO. L-003067 REV. NO. 1 PAGE NO. 7 OF 21 2.6.1. EAB and LPZ Doses at the Exclusion Area Boundary (EAB) and Low Population Zone (LPZ) for the FHA are based on the following formulas:

DosecEDE (rem) = Release (Curies) *1O(sec/m')

  • Breathing Rate (m 3/sec)
  • Inhalation DCF (remCEDE/Ci inhaled)

Q and Dose EoE (rem) = Release (Curies)

  • X (sec/m 3)
  • Submersion DCF (rem EoE - m 3/ Ci - sec)

Q and finally, DoseTEDE (rem)= DOSeCEDE (rem) + DoseEDE (rem) 2.6.2. Control Room The formulas used by RADTRAD,'by time increment, are:

DoseCEDE (rem) = Time Dependent CR Air Concentration (Ci/m 3)

  • Time Increment Duration (sec)
  • Breathing Rate (m3/sec)
  • Inhalation DCF (remCEDE/Ci inhaled)
  • Occupancy Factor and DoseEDE (rem) = Time Dependent CR Air Concentration (Ci/m 3)
  • Time Increment Duration (sec)
  • Submersion DCF (rem EDE - M3/ Ci - sec)
  • Occupancy Factor
  • CR Geometry Factor and finally, DoseTEDE (rem) = DoseCEDE (rem) + DoseEDE(rem) 2.7.

Acceptance Criteria Dose acceptance criteria are per 10CFR50.67 and RG 1.183 Table 6 guidance.

Table 2 lists the regulatory limits for accidental dose to 1) a control room operator, 2) a person at the EAB, and 3) a person at the LPZ boundary.

Table 2 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 B "Assumptions for Evaluating the Radiological Consequences of a Fuel Handling Accident" is provided by this analysis, as shown in the Conformance Matrix Table 3 below.

I CALCULATION NO. L-003067 REV. NO. 1 I PAGE NO. 8 OF 21 1

Acceptable assumptions regarding core inventory and the release of radionuclides from the fuel are provided in Regulatory Position 3 of this guide.

Conforms These assumptions are utilized; see Section 2 of this calculation 1.1 The number of fuel rods damaged during the accident should be based on a Conforms A conservative fuel conservative analysis that considers the most limiting case. This analysis should damage analysis consider parameters such as the weight of the dropped heavy load or the weight has been performed; of a dropped fuel assembly (plus any attached handling grapples), the height of see Section 2.1 and the drop, and the compression, torsion, and shear stresses on the irradiated fuel Attachment E of this rods. Damage to adjacent fuel assemblies, if applicable (e.g., events over the calculation.

reactor vessel), should be considered.

1.2 The fission product release from the breached fuel is based on Regulatory Conforms These assumptions Position 3.2 of this guide and the estimate of the number of fuel rods breached.

are utilized; see All the gap activity in the damaged rods is assumed to be instantaneously Section 2.2 of this released. Radionuclides that should be considered include xenons, kryptons, calculation.

halogens, cesiums, and rubidiums.

1.3 The chemical form of radioiodine released from the fuel to the spent fuel pool Conforms All iodine added to should be assumed to be 95% cesium iodide (CsI), 4.85 percent elemental iodine, pool is assumed to and 0.15 percent organic iodide. The CsI released from the fuel is assumed to dissociate.

completely dissociate in the pool water. Because of the low pH of the pool water, the iodine re-evolves as elemental iodine. This is assumed to occur instantaneously. The'NRC staff will consider, on a case-by-case basis, justifiable mechanistic treatment of the iodine release from the pool.

2 If the depth of water above the damaged fuel is 23 feet or greater, the Conforms; The decontamination factors for the elemental and organic species are 500 and 1, however, the more decontamination respectively, giving an overall effective decontamination factor of 200 (i.e., 99.5%

conservative factor was of the total iodine released from the damaged rods is retained by the water). This decontamination determined in a difference in decontamination factors for elemental (99.85%) and organic iodine factor (DF) of more conservative (0.15%) species results in the iodine above the water being composed of 57%

285.29 for manner than elemental and 43% organic species. If the depth of water is not 23 feet, the elemental iodine is prescribed in RG

I CALCULATION NO. L-003067 I REV. NO. 1 I PAGE NO. 9 OF 21 1

decontamination factor will have to be determined on a case-by-case method.

used since it is the value that yields an overall effective DF of 200 for 23 feet of water when combined with the stated initial iodine fractions.

1.183, as described in Section 2.3 of this calculation.

3 The retention of noble gases in the water in the fuel pool or reactor cavity is Conforms These assumptions negligible (i.e., decontamination factor of 1). Particulate radionuclides are are utilized.

assumed to be retained by the water in the fuel pool or reactor cavity (i.e., infinite decontamination factor).

4.1 The radioactive material that escapes from the fuel pool to the fuel building is Conforms This assumption is assumed to be released to the environment over a 2-hour time period, utilized. No credit is taken for the SGTS filtration.

4.2 A reduction in the amount of radioactive material released from the fuel pool by Not Applicable No credit is taken for engineered safety feature (ESF) filter systems may be taken into account filtration from the provided these systems meet the guidance of Regulatory Guide 1.52 and Generic reactor building.

Letter 99-02. Delays in radiation detection, actuation of the ESF filtration system, or diversion of ventilation flow to the ESF filtration system should be determined and accounted for in the radioactivity release analyses.

4.3 The radioactivity release from the fuel pool should be assumed to be drawn into Not Applicable Two-hour release to the ESF filtration system without mixing or dilution in the fuel building. If mixing the environment is can be demonstrated, credit for mixing and dilution may be considered on a case-assumed, without by-case basis. This evaluation should consider the magnitude of the building mixing or dilution.

volume and exhaust rate, the potential for bypass to the environment, the location of exhaust plenums relative to the surface of the pool, recirculation ventilation systems, and internal walls and floors that impede stream flow between the surface of the pool and the exhaust plenums.

I CALCULATION NO. L-003067 I REV. NO. 1 I PAGE NO. 10 OF 21 5.1 If the containment is isolated during fuel handling operations, no radiological consequences need to be analyzed.

Not Applicable This is considered Not Applicable to Boiling Water Reactors.

Containment is assumed isolated, but a permissible secondary containment opening during fuel handling is identified in Section 7.

5.2 If the containment is open during fuel handling operations, but designed to automatically isolate in the event of a fuel handling accident, the release duration should be based on delays in radiation detection and completion of containment isolation. If it can be shown that containment isolation occurs before radioactivity is released to the environment, no radiological consequences need to be analyzed.

Not Applicable This is considered Not Applicable to Boiling Water Reactors.

Containment is assumed isolated, but a permissible secondary containment opening during fuel handling is identified in Section 7.

5.3 If the containment is open during fuel handling operations (e.g., personnel air lock Conforms This 2-hour release or equipment hatch is open), the radioactive material that escapes from the assumption is reactor cavity pool to the containment is released to the environment over~a 2-utilized.

hour time period.

5.4 A reduction in the amount of radioactive material released from the containment Not Applicable No credit is taken for I by ESF filter systems may be taken into account provided that these systems filtration of release

I CALCULATION NO. L-003067 I REV. NO. 1 I PAGE NO. 11 OF 21 1

meet the guidance of Regulatory Guide 1.52 and Generic Letter 99-02. Delays in radiation detection, actuation of the ESF filtration system, or diversion of ventilation flow to the ESF filtration system should be determined and accounted for in the radioactivity release analyses.

from the reactor building.

5.5 Credit for dilution or mixing of the activity released from the reactor cavity by Not Applicable No credit is taken for natural or forced convection inside the containment may be considered on a dilution or mixing of case-by-case basis. Such credit is generally limited to 50% of the containment the activity released free volume. This evaluation should consider the magnitude of the containment from the reactor volume and exhaust rate, the potential for bypass to the environment, the location cavity.

of exhaust plenums relative to the surface of the reactor cavity, recirculation A 2-hour release ventilation systems, and internal walls and floors that impede stream flow assumption is between the surface of the reactor cavity and the exhaust plenums.

utilized.

R1I I CALCULATION NO. L-003067 I REV. NO. 1 IPAGE NO. 12 OF21

3.

ASSUMPTIONS Assumptions and bounding analyzed conditions regarding the fuel handling accident scenarios are provided below, based on Ref. 13 and other bases as stated.

1. Per Ref. 13, movement of recently irradiated fuel will not occur less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the fuel has occupied a critical reactor core, establishing a basis for a definition of "recently irradiated fuel".
2. Per Ref. 13, fuel bundle peak burnup will not exceed the RG 1.183 footnotes 10 and 11 limit of 62 GWD/MTU.
3. Per Ref. 13, for fuel exceeding a 54 GWD/MTU burnup, the maximum linear heat generation rate will not exceed the RG 1.183 footnote 11 limit of 6.3 kW/ft rod average power.
4. Per Ref. 13, the design basis GESTAR II bounding fuel damage assessment scenario associated with a drop over the reactor core is used. For this event the RG 1.183 DF value of 200 is conservative.
5. Spent fuel source terms are based on reactor core source terms as discussed in Attachment A.
6. Per Ref. 13. the damaged fuel is assumed to have operated at a radial peaking factor of 1.7.
7. Activity reaching the refuel floor airspace will essentially all be exhausted within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> (per Ref. 2) by using an artificially high exhaust rate. This also provides an allowance for uneven mixing in the refuel floor airspace.
8. Based on the LSCS plant configuration and an intact secondary containment, with the upper portion (above the refueling floor) of the Reactor Building facing the Auxiliary Building rooftop control room air intakes to the west of the Reactor Building, the bounding release pathway for the fuel handling accident release is considered a diffuse area source from the west Reactor Building wall. The actual release pathway, not credited, would be through the SGTS to the elevated stack.

R1i

9. In the event secondary containment integrity is not maintained, the following three pathways have been considered and evaluated in Section 7 (per ref. 6, any secondary containment openings not shown to be acceptable in this Design Analysis will not be permitted prior to a determination of their acceptability):
1. an assumed opening on the Reactor Building roof near the control room air intakes, RI
2.

an open truck bay door on the east side of the Reactor Building at grade level (for this pathway, a release from the refueling floor down the open hatchway to the grade level door is assumed),

R1 I

3. Open Integrated Leak Rate Test penetrations (MK-1RB-782 and MK-1RB-786) on the east side of the Reactor Building at grade level.

All of these represent conservative potential release pathways which would be closed rapidly in the event of an FHA per RG 1.183, Appendix B, footnote 3. However, the closure is not credited as mitigating the design basis release.

CALCULATION NO. L-003067 REV. NO. 1 PAGE NO. 13 OF 21 The possibility of a release of significance through openings into the adjacent Auxiliary Building or further to the Turbine Building is not considered credible. Released activity would have to be conveyed: (1) from the Reactor Building refuel floor atmosphere, (2) down the open hatchway to the lower floors with connections to the adjacent Auxiliary Building, (3) around the drywell and other obstructions to the opposite Reactor Building wall where these connections are located, (4) through the closed openings (such as double doors) to the adjacent buildings, and (5) from there to the outside either through the adjacent building's external openings or their ventilation systems. Such combinations of potential pathways are not considered credible as a mechanism for release to the environment.

All accesses pathways between the Reactor Building and the Auxiliary Building will be maintained closed during fuel movement, except for normal ingress/egress limited to that provided by a single open door.

10. No credit is taken for the operation of the CRAF System during the FHA. A worst case of 30,000 cfm (exceeding purge flow by - 14%) is used to conservatively minimize the time required for Control room radioactivity concentrations to approach those outside of the control room.

R 1 I R 1I CALCULATION NO. L-003067 REV. NO. 1 PAGE NO. 14 OF21

4.

DESIGN INPUT The design inputs used for this calculation are summarized in the following table:

TABLE 4 Parameters Applicable to AST Fuel Handling Accident Dose Calculations for LaSalle Generating Station (Note: all time periods indicated are from the initiation of the FHA)

Analysks rSuc oueis l~aimanteiiir Method.

AS 11tcSoreDouet Reactor Power 3559 MWth, including a 2%

Ref. 13 uncertainty Fuel Assembly Configuration 10xl0 in a 87.33 fuel pin Ref. 10, 11, 12 and 13 and properties bundle and 172 pins damaged Radial Peaking Factor 1.7 Ref. 13 Allowable Fuel Burnup and Limited to RG 1.183 footnote Table 3 of RG 1.183 and Ref.

non-LOCA gap fractions 10 and 11 value 13 FHA Radionuclide Inventory From Attachment A of this See Attachment A Calc. for the 60 isotopes forming the standard RADTRAD library, with decay to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> per Ref. 1 and 13, and Technical Specification 3 /

4.9.4, for fuel not recently irradiated Gap activities per Table 3 of RG 1.183.

Underwater Decontamination Noble Gases: 1 RG 1.183 Factor Particulate (cesium and RG 1.183 rubidium): infinity Iodine: 200, conservative RG 1.183 value for the limiting case of a (See Attachment E) drop over the reactor well.

Dose Conversion Factors EPA Federal Guidance Reports Ref. 4 and 5 11 and 12 Offsite Dose Limit 6.3 rem TEDE Table 6 of RG 1.183 Control Room Dose Limit 5 rem TEDE for the duration 10CFR50.67 of the accident Secondary Containment Not credited This analysis, in conformance Automatic Isolation and with RG 1.1 83 Filtration Mitigation by CRAF System Not credited This analysis, in conformance with RG 1.183 Bounding Control Room 30,000 cfm, or 14% above the Ref. 21 Fresh Air Intake purge flow rate of 26,340 cfmn.

CALCULATION NO. L-003067 REV. NO. 1 PAGE NO. 15 OF 21 Analysis jAST Value Source Documents Parameter or Method Control Room Volume Volume 117,400 ft3. (The Ref. 13 value of 117,500 ft3 is used in this analysis, as the difference is negligible.)

Reactor Building Normal Artificially set at an air change Conservative value for Ventilation rate rate of 0.1/min. This evacuates calculation

[Il ie(

2 hrs x 60 min/hr x 0.1/min)]

99.9994% of all activity within 2-hours.

CR Release Point Basis Sheet metal wall faces the CR Ref. 7 intake, so a worst case "diffuse area" source to the closest (south) CR intake is assumed with relaxed* Secondary Containment requirements.

Dispersion Factor 1.67E-03 sec/m 3 Ground Ref. 7 0 - 2 hr value (doses calculated through 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />)

EAB Release Point Basis and Plant Vent Stack, treated as a Distance to EAB ground level release with relaxed* Secondary Containment requirements.

Distance to EAB = 423 meters Ref. 13 Dispersion Factors 5.40E-04 sec/m 3 Ground Ref. 7 0 - 2 hr value LPZ Release Point Basis and Plant Vent Stack, treated as a Distance to LPZ ground level release with relaxed* Secondary Containment requirements.

Distance to LPZ 6400 meters Ref. 13 Dispersion Factors 2.26E-05 sec/m 3 Ground Ref. 7 0 - 2 hr value

  • Regarding immediate suspension of movement of irradiated fuel assemblies for fuel decayed at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

R1I R1I R1

[CALCULATION NO. L-003067 I REV. NO. 1 I PAGE NO. 16 OF 21

5.

REFERENCES

1.

LaSalle County Station Units 1 & 2, UFSAR, Revision 17, April 2008.

2.

Regulatory Guide 1.183, "Alternative Radiological Source Terms For Evaluating Design Basis Accidents At Nuclear Power Reactors", Rev. 0, July 2000.

3.

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.

4.

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

1988.

5.

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

6.

ATI/AR # 773640.

7.

LSCS Design Analysis L-003063, "Alternative Source Term Onsite and Offsite X/Q Values",

Rev. 1.

8.

G. Burley, "Evaluation of Fission Product Release and Transport", Staff Technical Paper, 1971.

9.

NEDC-32868P, "GEI4 Compliance With Amendment 22 of NEDE-24011-P-A (GESTAR II)", Rev. 1, September 2000.

10.

NEDE-31152P, "General Electric Fuel Bundle Designs" February 1993.

11.

NEDE-24011-P-A-14-US, General Electric Standard Application for Reactor Fuel, Licensing Topical Report, June 2000.

12.

Letter dated June 2, 2000 from J. Baumgartner, GNF Fuel Project Manager, to J.

Carmody, Exelon.

13.

SEAG 08-000075, Rev. 0, "LaSalle County Station Transmittal of Design Information (TODI) of Parameters and Source of Reference for AST Analysis".

14.

EMF-96-171(P), Rev. 2, "LaSalle Fuel Handling Accident for Atrium-9B Fuel".

15.

EMF-2679(P), Rev. 0, "LaSalle Fuel Handling Accident for Atrium-1 0 Fuel".

16.

LaSalle Drawing M-17, Rev. M.,"General Arrangement Section 'E-E' & 'F-F"'.

CALCULATION NO. L-003067 REV. NO. 1 PAGE NO. 17 OF 21

17.

Deleted.

18.

Deleted.

19.

GE Drawing 107E1592, Rev. 1, "Fuel Bundle".

20.

L-003128, Rev. 0, "LaSalle Source Terms for Use in Alternative Source Terms".

21.

LaSalle Drawing P&ID M-1443, Rev. R, "Control Room Air Conditioning System".

I CALCULATION NO. L-003067 I REV. NO. I I PAGE NO. 18 OF 21

6.

CALCULATIONS This calculation evaluates the radiological dose to an operator in the Control Room and a person at the EAB and LPZ locations following a design basis FHA involving irradiated fuel that has decayed to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after occupying a critical reactor core shutdown. Analyses are performed without certain safety features that have been historically credited in LaSalle FHA analyses, in order to determine that the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> decay time is sufficient to not credit those features to meet regulatory limits. This analysis uses Alternative Source Term assumptions per guidance in RG 1.183.

The RADTRAD v. 3.03 computer code was used for this calculation to determine FHA doses at the three dose points cited in RG 1.183; the Exclusion Area Boundary (EAB), Low Population Zone (LPZ), and the Control Room.

RADTRAD (Ref. 3) is a simplified model of RADionuclide Transport 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.

The technical basis for the RADTRAD code is documented in Section 2 of NUREG/CR-6604. The methodologies significant to this analysis are the dose consequence analysis (NUREG Section 2.3) and the Radioactive Decay Calculations (NUREG Section 2.4).

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.

See Section 2 and subsequent sections for descriptions of the calculational simulations detailed in the tables that follow.

6.1.1. RADTRAD Run Compartment Information RADTPAD Compar:ipt nts Compartment 1

2 3

Number Name Reactor Building Environment Control Room Type Other Environment Control Room Volume (ft3) 1 0

117,500 Source Term Frct 1.000 0.000 0.000 Fraction:

Compartment Used to represent the No credited removal Features ventilated radionuclide Environment mechanisms, including existing release region above the pool filtered recirc l

system.

and water during refueling.

Comments Actual volume is irrelevant, as Control Room doses calculated the ventilated fraction is used.

without credit for CRAF system filtration features and with normal Control Room ventilation system in service at a worst-case outside air intake rate from Section 4

_above the purge flow rate.

CALCULATION NO. L-003067 REV. NO. 1 I PAGE NO. 19 OF 21 6.1.2.

RADTRAD Run Transfer Pathway Information

~ RADTRAD Transfer, Pathway's Pathway Number 1

2 3

Name Reactor Building to (High Volume-Purge+

(Control Room Exhaust)

Environment Intake) Environment to Control Room to Control Room Environment From-To 1-2 2-3 3-2 Transfer Mechanism Filter Filter Filter Transfer Mechanism Filter Panel - Flow rate for 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Filter Panel - Flow rate for 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Filter Panel - Flow rate for 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period at 0.1 cfm period at 30,000 cfm.

period at 30,000 cfm.

Details Filter Efficiency - Efficiency is Filter Efficiency - Efficiency is Filter Efficiency - Efficiency is entered entered as 0.0% for all chemical entered as 0.0% for all chemical as 100.0% for all chemical forms of forms of iodine, for the accident forms of iodine, for the accident iodine, for all time periods.

duration, duration.

Comments Vented at an artificially high 0.1/min CRAF System intake and This is the exhaust from the control air change rate without credit for recirculation filtration not credited, room to the environment; the filtration SGTS filtration.

prevents a double counting of the iodine release. Although RADTRAD 3.03 documentation indicates that this effect has been eliminated, this was still done for completeness.

6.1.3.

RADTRAD Run Dose Location Information

.. RADTRAD Dose Locations Dose Location 2

3 Number Name Exclusion Area Boundary Low Population Zone (LPZ)

Control Room (EAB)

Lo_

ouato_

oe_ LZonrlRo In Compartment Environment (2)

Environment (2)

Control Room (3)

Breathing Rate (m3/sec) 3.5E-04 3.5E-04 3.5E-04 Occupancy 0-24-hour Period:

1.0 Fractions Dispersion Ground Level Release:

Ground Level Release:

Factors 5.40E-04 2.26E-05 Release:

(sec/m3) 1.67E-03 Comments Ground level release dispersion Ground level release dispersion Worst case release dispersion factors factors are used from Table 4 factors are used from Table 4 are used from Table 4 because the because the analyses do not credit because the analyses do not credit analyses do not credit SGTS filtration or SGTS filtration or Secondary SGTS filtration or Secondary Secondary Containment.

Containment.

Containment.

Release is for two hours. Control Room Breathing Rate and Occupancy Breathing Rate and Occupancy dose calculated until effectively all Fraction per Ref. 2.

Fraction per Ref. 2.

activity has been exhausted.

Breathing Rate and Occupancy Fraction per Ref. 2.

CALCULATION NO. L-003067 I REV. NO. 1 PAGE NO. 20 OF 21 6.1.4. RADTRAD Run Source Term & Dose Conversion Factor Information RADTRAD Source Term & Dose Conversion Factors Core Power 3559 MWth; 764 assemblies; 172 damaged rods with 87.33 rods/assembly; and Radial Peaking Factor of 1.7 =

3559 MWth

  • 1.7*172/(764*87.33) =See Section 2.1.

15.597 MWth in damaged fuel Nuclide Inventory Attachment A ORIGEN calculated core inventory for 60 MACCS isotopes. Inventory, calculated at DBA power shown above, and input into the FHA NIF file with adjustments See Attachments A and C.

for Kr-85 and 1-131 (Attachment C, with the Kr-85 Attachment A value multiplied by 2.0 and the 1-131 Attachment A value multiplied by 1.6, as indicated).

Release Fractions User defined Release Fractions to calculate

& Timing as effectively 100% instantaneous, to ensure that release timing is not delayed, consistent with RG 1.183 guidance. Fractions shown in RFT file includes gap activity fractions, combined for Iodine with the applicable See Section 2.3 and Attachment D.

decontamination factor (200 for a minimum water coverage depth of 23 feet, so the Iodine RFT value is 0.05 Halogen Gap Release Fraction divided by 200 = 2.50 E-04).

Dose Conversion RADTRAD Library of FGR 11 & 12 values for Refs. 4 and 5.

Factors 60 MACCS isotopes.

Decay &

Daughter Decay and daughter products are considered.

Products Iodine Chemical Aerosol:

0.0000 User defined iodine chemical fractions, Fractions Elemental:

0.9700 consistent with RG 1.183 guidance.

Organic:

0.0300 Note that no filters are credited so chemical I_

fractions are not important.

WGI has pre-qualified RADTRAD for application to perform such calculations, as documented in the Computer Disclosure Sheet F-1 of Attachment F.

R1I R1 R1i CALCULATION NO. L-003067 REV. NO. 1 PAGE NO. 21 OF 21

7.

RESULTS AND CONCLUSIONS The RADTRAD code was used to examine the effect of the alternative source term release on offsite and CR doses. Shown in Table 5 are the results, as well as the dose acceptance criteria.

Table 5 RADTRAD Analysis Results and Comparisons to the Acceptance Criteria Limits I6ocati~n Limits RADT RADAResuIts rem.TEDE)

EAB 6.3 1.50 LPZ 6.3 0.063 CR 5.0 3.35 These results indicate that the calculated consequences of a design basis FHA at or after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> since the fuel has occupied part of a critical reactor core will be within regulatory limits.

Since the above analyses are performed with no SGTS filtration, and also no CRAF System filtration, these results indicate that immediate suspension of movement of irradiated fuel assemblies in the secondary containment would not be necessary if Technical Specification Limiting Conditions of Operation TS 3.6.4.1A, TS 3.6.4.2A, TS 3.6.4.3A, TS 3.7.4.A, or TS 3.7.5.A, for example, were to occur.

Ref. 7 derives CR dispersion factors for potential pathways of assumed intentional opening of two doors in series during the FHA for the Reactor Building access to the Auxiliary Building roof, or for the Reactor Building Integrated Leak Rate Test (ILRT) penetrations near column 10 four feet above grade or truck bay doors between columns 14 and 15 at grade level. Table 6 below compares the CR doses from use of these dispersion factors to that used for the results above (as determined using the ratio of the pathway dispersion factor to the diffuse area release dispersion factor times the diffuse area release CR dose). The table indicates the opening of the two doors in series of the Reactor Building access to the Auxiliary Building roof cannot be allowed during fuel handling, but the temporary opening of the truck bay and / or an ILRT penetration can be allowed (e.g., during a dual unit shutdown). Also shown are the CR doses resulting from use of the Ref. 7 stack release atmospheric dispersion factors at stack discharge elevation and ground level to the closest CR intake (an artificial worse case), consistent with secondary containment integrity without filtration or any assumed diffuse area release.

Table 6 CR Dose Results Comparisons for Assumed Release Pathways and Their Acceptability Corresponding CR Dose Acceptability of the AsumdCRAtjmospheric Results Assumed pathway AsmdRelease Pathway Diprin (e

ihRespect to the Dispersion (rem Withrs~o Factor (seclM3 TEDE CR Dose Limit Worst Case Diffuse Area Release 1.67E-03 3.35 Yes Reactor Building Access to the Auxiliary Building Roof 7.15E-03 14.3 No Reactor Building Truck Bay Doors at Grade Level 5.71E-04 1.15 Yes Stack Discharge at Stack Elevation 1.19E-05 0.024 Yes Stack Discharge at Ground Elevation 6.84E-04 1.37 Yes Integrated Leak rate Test Reactor Building Penetration 9.08E-04 1.82 Yes Access to the Auxiliary Electric Room (AEER) and/or its contained Remote Shutdown Panel is not required in response to a FHA. As per Ref. 13, AEER occupancy is only required for the safety related action following a postulated LOCA of engaging fans that provide containment air mixing for combustible gas control. This mission and its dose consequences are evaluated for the AST LOCA.

ATTACHMENT A Source Terms (Following the Introduction section below, the following Source Terms derivation is extracted directly from Exelon Design Analysis No. L-003128, Rev. 0, as prepared by Robert Jaffa, reviewed by Moussa Mahgerefteh, and approved by Carlos Delahoz, all on 5/6/05).

Introduction The following calculation extract from Exelon Nuclear Design Analysis No. L-003128, Rev. 0, derives the isotopic inventory (core source terms) to be used as inputs for the LSCS AST Design Basis Accident analyses. The derivation uses an Exelon Nuclear controlled-version of ORIGEN 2.1, as noted.

RADTRAD uses a Nuclide Information File (NIF) file (as presented in Attachment C, for FHA) in units of Curies/MWt and multiplies this by the specified core power. The LSCS AST analyses utilize a Design Basis Accident power level of 3559 MWt., and 3559 MWt is used as the input specified core power for LSCS full core power analyses. The Curies/MWt utilized is the bounding isotopic core inventory, using the maximum quantity for each isotope as derived in the following Tables of either:

the 100 Effective Full Power Days (EFPD) calculated value (conservatively representing the Beginning of Cycle condition to ensure that all isotopes with half lives < 1 year are at equilibrium levels), or the calculated End of Cycle values, with full power operation conservatively assumed for the entirety (i.e., with no power coastdown) of a 711 EFPD cycle the bounding of the two equilibrium fuel cycles considered [either the Peach Bottom Unit 3 (PB-

3) cycle or the LaSalle Unit 1 (LS-1) cycle].

Default RADTRAD values are used for Co58 and Co60.

The resulting source term provides bounding doses consistent with Regulatory Guide 1.183, Section 3.1 for the applicable range of fuel enrichments and burnups. Peach Bottom AST sensitivity analyses documented in the following Ref. 3.6 confirmed that all Control Room (CR) and offsite doses for the assumed Loss of Coolant Accident are bounded by this "Base Case" source term approach of the first two bullets above, compared with consideration of alternatives such as the following:

Base Case except for higher enrichment with utilization of the maximum calculated Curies for each isotope of either the 100 EFPD or EOC values Base Case except with higher enrichment and artificially bounding full-power cycle length of 740 EFPD with utilization of the maximum calculated Curies for each isotope of either the 100 EFPD or EOC values Base Case except with artificially bounding full-power cycle length of 740 EFPD, but with utilization of only the EOC values of Curies for each isotope Base Case except with a one-year "short cycle" of lower enrichment and full-power cycle length of 351 EFPD with utilization of the maximum calculated Curies for each isotope of either the 100 EFPD or EOC values Calc. No. L-003067, Rev. 1, Attachment A, Page Al of A14

1.0 PURPOSE To calculate the isotopic core inventory for LaSalle Units 1 and 2 using the ORIGEN2.1 code based on reactor operation at 3489 MWt and an equilibrium 711 EFPD two-year cycle design.

Development of Core Source Terms based on ORIGEN2.1 complies with RG 1.183. The results of this analysis are to be used as input for dose calculations using the Alternative Source Term methodology.

2.0 RESULTS and CONCLUSIONS The bounding isotopic core inventory for a 711 EFPD two-year cycle design at LaSalle Units 1 and 2 is shown in terms of activity (Cuires) in Table 2. This bounding isotopic core inventory was determined for LaSalle's rated power level of 3489 MWt. This result was then contrasted (Table 3) with a source term similarly calculated for the Peach Bottom Station for a 711 EFPD two-year cycle design at 3514.9 MWt. These two source terms were normalized to Ci/MWt units as shown in Table 4.

The bounding source terms are the maximum for each isotope considered of the Beginning of Cycle (BOC, or 100 Effective Full Power Days) or the End of Cycle (EOC) values for the two equilibrium fuel cycles developed in the Source Document. The use of BOC-EOC worst cases assures that worst accident timing is considered which provides inherent calculation margin as shown in the PBAPS Letter to the NRC dated December 8, 2004 providing responses to NRC RAIs.

Note that the AREVA cycle was artificially extended from 680 EFPDs to 711 EFPDs. The PBAPS source terms reflect a higher enrichment and maximize the inventory of isotopes with higher yields from U-235 fissions. The extended burnup AREVA case maximizes the inventory of isotopes with higher yields from Pu-239 and Pu-241.

3.0 REFERENCES

3.1 RSIC Code Package CCC-371, "ORIGEN 2.1, Isotope Generation and Depletion Code Matrix Exponential Method," May 1999.

3.2 AREVA letter RJD:05:002 from C. Powers to R. DeMartino, Re: "LaSalle Unit 1 Long Term Fuel Cycle Analysis," dated March 11, 2005.

3.3 Memo from R. Jaffa to R. Tropasso, Re: "Installation Verification of ORIGEN2.1 on the IBM PC Platform," dated July 7, 2000.

3.4 ORNL/TM-11018, "Standard-and Extended-Bumup PWR and BWR Reactor Models for the ORIGEN2 Computer Code," S. Ludwig, J. Renier, December 1989.

3.5 PBAPS Design Analysis PM-1059, "Re-analysis of Fuel Handling Accident (FHA) Using Alternative Source Terms", Rev. 2, January 4, 2005 3.6 Exelon letter to USNRC dated December 8, 2004, RE: "Supplement to the Request for License Amendments Related to Application of Alternative Source Term, dated July 14, 2003" 4.0 ASSUMPTIONS Caic. No. L-003067, Rev. 1, Attachment A, Page A2 of A14

4.1 The ORIGEN2.1 code [Ref 3.1] was used to calculate isotopic activities based on the cycle design described in Reference 3.2. The ORIGEN2.1 code was run on an IBM-PC and was confirmed to be controlled per Ref. 3.3.

4.2 Batch-average enrichments and exposures from Ref. 3.2 were used to develop input to ORIGEN2. 1. This is equivalent to performing individual calculations for each sub-batch.

4.3 For fuel burned-in more than one cycle, ORIGEN2.1 runs ignored refueling outages.

This has no impact on short-lived isotopes which reach equilibrium concentrations shortly after cycle startup and has a conservative, albeit minimal, impact on long-lived isotopes which continually increase in concentration as a function of exposure; ignoring intermediate decay periods will increase the final concentrations.

4.4 For isotopic activities calculated at BOC, BOC is defined as 100 days into the cycle to ensure that all short-lived isotopes (half lives < 1 year) are at equilibrium levels.

4.5 Extending the LaSalle cycle lengths from 680 EFPD to 711 EFPD without raising initial enrichments is conservative as lower enrichments result in higher activities for a given specific power.

4.6 The equilibrium core design developed for LaSalle Unit 1 in Ref. 3.2 is representative for LaSalle Unit 2.

5.0 INPUTS 5.1 Batch-average burnup and batch size information for an equilibrium two-year cycle was obtained from Table 10 of Ref. 3.2 for fuel projected to be loaded in LS-1 Cycle 15 5.2 Batch-average uranium loading and initial enrichments for an equilibrium two-year cycle was obtained from Table 4 of Ref. 3.2 for fuel projected to be loaded in LS-1 Cycle 15 5.3 Cycle energy for an equilibrium two-year cycle was obtained from Table 9 of Ref. 3.2 for projected LS-1 Cycle 15 5.4 The ORIGEN2.1 cross-section library, BWRUE.LIB, is used in this calculation as this is most representative of current LaSalle two-year cycles.

The library is based on an "extended cycle" reactor model where fuel achieves 40 GWd/mtU bumup in four cycles.

5.5 Equilibrium cycle lengths from Ref. 3.2 are for a 680 EFPD cycle length. ORIGEN2 cases were run extending the cycle lengths to 711 EFPD to cover reasonable extensions of the equilibrium cycle and to allow for comparison to a similar analysis performed for Peach Bottom 3.

5.6 The rated power level used for determining exposure data for the equilibrium two-year cycle is 3489 MW.

6.0 METHOD OF ANALYSIS Calc. No. L-003067, Rev. 1, Attachment A, Page A3 of A14

The isotopic core inventory is a function of the reactor power level and the exposure of the fuel.

The LaSalle Unit 1 isotopic core inventory is calculated using the ORIGEN2.1 code [Ref. 3.1]. A 3489 MWt 680 EFPD two-year equilibrium cycle design [from Ref. 3.2] is used as the basis for this calculation 6.1 Identification of Computer Programs The batch file used to execute ORIGEN2.1 (DTSQA Product ID# EX0004724) for this calculation provide the paths and filenames of the executable program and libraries that were called. The batch file used is Isl ast.bat. The PC-based ORIGEN2.1 code used in this calculation was verified to be controlled by comparing the time/date/size stamp of the executable file to that documented in Ref. 3.3.

Volume in drive D is PER30290 Volume Serial Number is 1 1F7-OC 1F Directory of D:\\Origen2 1\\CODE ORIGEN2 EXE 1,267,348 06-10-96 l:09p ORIGEN2.EXE The time/date/size stamps of the library files used in this calculation were verified against those documented in Ref. 3.4.

Volume in drive D is PER30290 Volume Serial Number is 1 1F7-OC1F Directory of D:\\Origen2 1\\LIBS BWRUE LIB 173,676 08-01-91 2:10a DECAY LIB 278,636 08-01-91 2:10a GXUO2BRM LIB 167,526 08-01-91 2:10a Caic. No. L-003067, Rev. 1, Attachment A, Page A4 of A14

7.0 NUMERIC ANALYSIS Equilibrium Two-Year Cycle Isotopic Core Inventory An equilibrium two-year cycle design for LaSalle Unit 1 using the projected Cycle 15-17 designs as a reference cycle and a cycle length of 711 EFPD is used as the basis for the source term calculation. The resulting batch-average burnups for once-burned, twice-burned and thrice-burned fuel batches are shown below.

Avg. Bumup Avg. Enr.

per Cycle Power I Loading U235 wt.

U238 wt.

Oxygen wt.

Batch

  1. of FA (w/o U235)

(MWd/mtU)

(MW)

(MTU)

(gins)

(gins)

(gins) 1 188 3.872 20515.02 1010.5 33.5016 1,297,312.24 32,204,287.76 4,504,247.89 19330.68 952.2 8295.98 408.6 2

288 3.872 21060.15 1589.2 51.3216 1,987,371.94 49,334,228.06 6,900,124.43 19717.64 1487.8 3

288 3.872 21060.15 1589.2 51.3216 1,987,371.94 49,334,228.06 6,900,124.43 U Loading per Bundle =

178.2 kgU Ref. 3.2 Table 4 - average mass Avg. Enrich. per Bundle =

3.872 w/o U235 Ref. 3.2 Table 4 - average enr.

Nominal Cycle Energy =

2373 GWd Ref. 3.2 Table 9 Rated Thermal Power =

3489 MW Nominal Cycle Length =

680.14 EFPD (calculated from Nominal Cycle Energy / Reated Thermal Power)

The equilibrium cycle isotopic core inventory is calculated using ORIGEN2.1 and the BWR extended burmup cross-section library BWRUE. The input deck is Islast.inp and the batch file is Islast.bat.

The specific power for a batch in a given cycle is determined by multiplying the batch average burnup for that cycle by the batch loading and then dividing by the number of EFPD in the cycle. For example, the specific power for Batch 1 in its first cycle of operation is:

(20,515.02

  • 33.5016) / 680.14 = 1010.5 MW.

The grams of U235 and U238 for each batch were determined by the following formulas:

U235 (gms)= Batch loading * (Avg. Enr./100)

  • 106 U238 (gms) = Batch loading * (1 - Avg. Enr./100)
  • 106 The corresponding weight of oxygen in U0 2 pellets for each batch is:

0 (gms) = Total batch U weight (gin U) / 238 (gin U/gm atom U)

  • 2 (gin atom 0/gm atom U)
  • 15.9994 gm 0/gm atom 0 Caic. No. L-003067, Rev. 1, Attachment A, Page A5 of A14

The ORIGEN2.1 input deck is set up to deplete each fuel batch and write the 100 EFPD and EOC results to temporary storage vectors. Once all batches have been depleted, the results from the temporary vectors are combined to give the results for the entire core. The ORIGEN2.1 core inventory activity for.the equilibrium two-year cycle at 100 EFPD (BOC) and EOC is shown below in Table 1. The maximum of the 100 EFPD and EOC values for each isotope are selected to generate the bounding isotopic core inventory activity as shown in Table 2.

The bounding LaSalle source term shown in Table 2 is compared against the bounding Peach Bottom source term from Ref. 3.5 in Table 3. As expected, results are very similar due to the similarity of the core designs. Table 4 shows a comparison of the bounding LaSalle source term vs. the bounding Peach Bottom source term in terms of normalized activity (Ci/MW). The Peach Bottom and LaSalle bounding source terms in Table 3 are divided by core thermal powers of 3514.9 MW and 3489 MW, respectively.

Calc. No. L-003067, Rev. 1, Attachment A, Page A6 of A14

Table 1 ORIGEN2.1 Isotopic Activity Results for LaSalle Unit I 100 EFPD EOC 100 EFPD EOC Isotope T

UOT

[

TI KR 83M 1.313E+07 1.137E+07 BR 84 2.352E+07 1.958E+07 BR 85 2.861E+07 2.348E+07 KR 85 8.297E+05 1.331E+06 KR 85M 2.895E+07 2.380E+07 RB 86 1.032E+05 2.236E+05 KR 87 5.684E+07 4.552E+07 KR 88 8.018E+07 6.400E+07 RB 88 8.118E+07 6.507E+07 SR 89 9.713E+07 8.630E+07 SR 90 6.545E+06 1.066E+07 Y90 6.692E+06 1.099E+07 SR 91 1.323E+08 1.083E+08 Y 91 1.196E+08 1.118E+08 SR 92 1.399E+08 1.179E+08 Y 92 1.404E+08 1.185E+08 Y93 1.578E+08 1.379E+08 ZR 95 1.505E+08 1.557E+08 NB 95 1.381E+08 1.565E+08 ZR 97 1.624E+08 1.562E+08 MO 99 1.755E+08 1.770E+08 TC 99M 1.536E+08 1.549E+08 RU103 1.227E+08 1.480E+08 RU105 7.608E+07 1.035E+08 RH105 7.254E+07 9.755E+07 RU106 3.872E+07 6.141 E+07 SB127 8.573E+06 1.027E+07 TE127 8.407E+06 1.018E+07 TE127M 1.043E+06 1.364E+06 SB129 2.723E+07 3.044E+07 TE129 2.670E+07 2.995E+07 TE129M 3.861E+06 4.461E+06 1129 2.657E+00 4.704E+00 TE131M 1.263E+07 1.358E+07 1131 9.078E+07 9.402E+07 Isotope

( iT)

(I L )

XE131M 1.008E+06 1.051E+06 T E132 1.313E+08 1.33TE+08 1132 1.329E+08 1.357E+08 1133 1.938E+08 1.909E+08 XE133 1.889E+08 1.914E+08 XE133M 5.914E+06 5.968E+06 1134 2.149E+08 2.098E+08 CS134 1.293E+07 2.482E+07 1135 1.810E+08 1.791E+08 XE135 7.543E+07 6.836E+07 XE135M 3.612E+07 3.758E+07 CS136 3.316E+06 6.898E+06 CS137 8.989E+06 1.544E+07 BA137M 8.519E+06 1.462E+07 XE138 1.664E+08 1.569E+08 CS138 1.825E+08 1.740E+08 BA139 1.772E+08 1.701E+08 BA140 1.707E+08 1.643E+08 LA140 1.747E+08 1.703E+08 LA141 1.618E+08 1.546E+08 CE141 1.565E+08 1.558E+08 LA142 1.580E+08 1.492E+08 CE143 1.542E+08 1.429E+08 PR143 1.496E+08 1.396E+08 CE144 9.895E+07 1.246E+08 ND147 6.402E+07 6.256E+07 NP239 1.650E+09 1.933E+09 PU238 2.521E+05 6.011E+05 PU239 3.057E+04 4.212E+04 PU240 2.842E+04 4.565E+04 PU241 1.327E+07 2.183E+07 AM241 1.511 E+04 3.418E+04 CM242 3.346E+06 7.894E+06 CM244 2.858E+05 9.007E+05 Calc. No. L-003067, Rev. 1, Attachment A, Page A7 of A14

Table 2 Bounding Isotopic Core Inventory LaSalle Unit 1 Isotooic Isotooic Activity Isotope (Ci)

KR 83M 1-3131F+07 Activity Isotope (Ci)

XFI31M 1 0.51F+06 BR 84 BR 85 KR 85 KR 85M RB 86 KR 87 KR 88 RB 88 SR 89 SR 90 Y 90 SR 91 Y91 SR 92 Y 92 Y93 ZR 95 NB 95 ZR 97 MO 99 TC 99M RU103 RU105 RH105 RU106 SB127 TE127 TE127M SB129 TE129 TE129M 1129 TE131M 1131 2.352E+07 2.861 E+07 1.331 E+06 2.895E+07 2.236E+05 5.684E+07 8.018E+07 8.118E+07 9.713E+07 1.066E+07 1.099E+07 1.323E+08 1.196E+08 1.399E+08 1.404E+08 1.578E+08 1.557E+08 1.565E+08 1.624E+08 1.770E+08 1.549E+08 1.480E+08 1.035E+08 9.755E+07 6.141E+07 1.027E+07 1.018E+07 1.364E+06 3.044E+07 2.995E+07 4.461 E+06 4.704E+00 1.358E+07 9.402E+07 TE132 1132 1133 XE133 XE133M 1134 CS134 1135 XE135 XE135M CS136 CS137 BA137M XE138 CS138 BA139 BA140 LA140 LA141 CE141 LA142 CE143 PR143 CE144 ND147 NP239 PU238 PU239 PU240 PU241 AM241 CM242 CM244 1.336E+08 1.357E+08 1.938E+08 1.914E+08 5.968E+06 2.149E+08 2.482E+07 1.810E+08 7.543E+07 3.758E+07 6.898E+06 1.544E+07 1.462E+07 1.664E+08 1.825E+08 1.772E+08 1.707E+08 1.747E+08 1.618E+08 1.565E+08 1.580E+08 1.542E+08 1.496E+08 1.246E+08 6.402E+07 1.933E+09 6.011 E+05 4.212E+04 4.565E+04 2.183E+07 3.418E+04 7.894E+06 9.007E+05 Caic. No. L-003067, Rev. 1 Attachment A, Page A8 of Al 4

Table 3 Bounding Isotopic Core Inventory Total Activity Comparison PB-3 Bounding LS-1 Bounding PB-3 Bounding LS-1 Bounding Isotopic Isotopic (Lb - PB)/PB Activity Activity Change Isotope (Ci)

(Ci)

KR 83M 1.324E+07 1.313E+07

-0.83 BR 84 BR 85 KR 85 KR 85M RB 86 KR 87 KR 88 RB 88 SR 89 SR 90 Y 90 SR 91 Y 91 SR 92 Y 92 Y93 ZR 95 NB 95 ZR 97 MO 99 TC 99M RU103 RU105 RH105 RU 106 SB127 TE127 TE127M SB129 TE129 TE129M 1129 TE131M 1131 Boldina is 2.373E+07 2.888E+07 1.387E+06 2.922E+07 2.291E+05 5.739E+07 8.096E+07 8.197E+07 9.836E+07 1.117E+07 1.150E+07 1.336E+08 1.212E+08 1.412E+08 1.416E+08 1.591 E+08 1.578E+08 1.586E+08 1.637E+08 1.785E+08 1.563E+08 1.477E+08 1.022E+08 9.673E+07 6.081 E+07 1.018E+07 1.010E+07 1.355E+06 3.036E+07 2.988E+07 4.453E+06 4.816E+00 1.360E+07 9.444E+07 ased if the 2.3b2E+U0 2.861 E+07 1.331 E+06 2.895E+07 2.236E+05 5.684E+07 8.018E+07 8.118E+07 9.713E+07 1.066E+07 1.099E+07 1.323E+08 1.196E+08 1.399E+08 1.404E+08 1.578E+08 1.557E+08 1.565E+08 1.624E+08 1.770E+08 1.549E+08 1.480E+08 1.035E+08 9.755E+07 6.141 E+07 1.027E+07 1.018E+07 1.364E+06 3.044E+07 2.995E+07 4.461 E+06 4.704E+00 1.358E+07 9.402E+07

-0.88

-0.93

-4.04

-0.92

-2.40

-0.96

-0.96

-0.96

-1.25

-4.57

-4.43

-0.97

-1.32

-0.92

-0.85

-0.82

-1.33

-1.32

-0.79

-0.84

-0.90 0.20 1.27 0.85 0.99 0.88 0.79 0.66 0.26 0.23 0.18

-2.33

-0.15

-0.44 Isotopic Isotopic

-B)/B Activity Activity Change Isotope (Ci)

(Ci)

XE131M 1.056E+06 1.051 E+06

-0.47 1E132 1.343E+08 1.336E+08

-0.52 1132 1.364E+08 1.357E+08

-0.51 1133 1.953E+08 1.938E+08

-0.77 XE133 1.930E+08 1.914E+08

-0.83 XE133M 6.007E+06 5.968E+06

-0.65 1134 2.167E+08 2.149E+08

-0.83 CS134 2.559E+07 2.482E+07

-3.01 1135 1.825E+08 1.810E+08

-0.82 XE135 7.832E+07 7.543E+07

-3.69 XE135M 3.773E+07 3.758E+07

-0.40 CS136 7.123E+06 6.898E+06

-3.16 CS137 1.595E+07 1.544E+07

-3.20 BA137M 1.510E+07 1.462E+07

-3.18 XE138 1.679E+08 1.664E+08

-0.89 CS138 1.841E+08 1.825E+08

-0.87 BA139 1.787E+08 1.772E+08

-0.84 BA140 1.721 E+08 1.707E+08

-0.81 LA140 1.764E+08 1.747E+08

-0.96 LA141 1.631E+08 1.618E+08

-0.80 CE141 1.579E+08 1.565E+08

-0.89 LA142 1.593E+08 1.580E+08

-0.82 CE143 1.556E+08 1.542E+08

-0.90 PR143 1.509E+08 1.496E+08

-0.86 CE144 1.264E+08 1.246E+08

-1.42 ND147 6.459E+07 6.402E+07

-0.88 NP239 1.897E+09 1.933E+09 1.90 PU238 6.312E+05 6.011 E+05

-4.77 PU239 4.218E+04 4.212E+04

-0.14 PU240 4.526E+04 4.565E+04 0.86 PU241 2.173E+07 2.183E+07 0.46 AM241 3.349E+04 3.418E+04 2.06 CM242 8.393E+06 7.894E+06

-5.95 CM244 9.147E+05 9.007E+05

-1.53 Last column is positive, implying that LS bounds.

Calc. No. L-003067, Rev. 1, Attachment A, Page A9 of A14

Table 4 Bounding Isotopic Core Inventory Normalized Activity Comparison PB-3 Bounding LS-1 Bounding Isotopic Isotopic (LS - PB)/PB Activity Activity Change Isotope (Ci/MW)

(Ci/MW)

KR 83M 3.767E+03 3.763E+03

-0.09 BR 84 6.751E+03 6.741E+03

-0.15 BR 85 8.216E+03 8.200E+03

-0.20 KR 85 3.946E+02 3.815E+02

-3.33 KR 85M 8.313E+03 8.298E+03

-0.19 RB 86 6.518E+01 6.409E+01

-1.68 KR 87 1.633E+04 1.629E+04

-0.22 KR 88 2.303E+04 2.298E+04

-0.23 RB 88 2.332E+04 2.327E+04

-0.23 SR 89 2.798E+04 2.784E+04

-0.52 SR 90 3.178E+03 3.055E+03

-3.86 Y 90 3.272E+03 3.150E+03

-3.73 SR 91 3.801E+04 3.792E+04

-0.24 Y 91 3.448E+04 3.428E+04

-0.59 SR 92 4.017E+04 4.010E+04

-0.19 Y 92 4.029E+04 4.024E+04

-0.11 Y 93 4.526E+04 4.523E+04

-0.08 ZR 95 4.489E+04 4.463E+04

-0.60 NB 95 4.512E+04 4.486E+04

-0.59 ZR 97 4.657E+04 4.655E+04

-0.06 MO 99 5.078E+04 5.073E+04

-0.10 TC 99M 4.447E+04 4.440E+04

-0.16 RU103 4.202E+04 4.242E+04 0.95 RU105 2.908E+04 2.966E+04 2.02 RH105 2.752E+04 2.796E+04 1.60 RU106 1.730E+04 1.760E+04 1.74 SB127 2.896E+03 2.944E+03 1.63 TE127 2.873E+03 2.918E+03 1.54 TE127M 3.855E+02 3.909E+02 1.41 SB129 8.638E+03 8.725E+03 1.01 TE129 8.501 E+03 8.584E+03 0.98 TE129M 1.267E+03 1.279E+03 0.92 1129 1.370E-03

.1.348E-03

-1.60 TE131M 3.869E+03 3.892E+03 0.59 1131 2.687E+04 2.695E+04 0.29 PB-3 Bounding LS-1 Bounding Isotopic Isotopic (LS - PB)/PB Activity Activity Change Isotope (Ci/MW)

(Ci/MW)

XE131M 3.004E+02 3.012E+02 0.27 TE132 3.821 E+04 3.829E+04 0.22 1132 3.881 E+04 3.889E+04 0.23 1133 5.556E+04 5.555E+04

-0.03 XE133 5.491 E+04 5.486E+04

-0.09 XE133M 1.709E+03 1.711 E+03 0.09 1134 6.165E+04 6.159E+04

-0.09 CS134 7.280E+03 7.114E+03

-2.29 1135 5.192E+04 5.188E+04

-0.09 XE135 2.228E+04 2.162E+04

-2.98 XE135M 1.073E+04 1.077E+04 0.34 CS136 2.027E+03 1.977E+03

-2.44 CS137 4.538E+03 4.425E+03

-2.48 BA137M 4.296E+03 4.190E+03

-2.46 XE138 4.777E+04 4.769E+04

-0.16 CS138 5.238E+04 5.231 E+04

-0.13 BA139 5.084E+04 5.079E+04

-0.10 BA140 4.896E+04 4.893E+04

-0.08 LA140 5.019E+04 5.007E+04

-0.23 LA141 4.640E+04 4.637E+04

-0.06 CE141 4.492E+04 4.486E+04

-0.15 LA142 4.532E+04 4.529E+04

-0.08 CE143 4.427E+04 4.420E+04

-0.16 PR143 4.293E+04 4.288E+04

-0.13 CE144 3.596E+04 3.571 E+04

-0.69 ND147 1.838E+04 1.835E+04

-0.15 NP239 5.397E+05 5.540E+05 2.65 PU238 1.796E+02 1.723E+02

-4.06 PU239 1.200E+01 1.207E+01 0.60 PU240 1.288E+01 1.308E+01 1.61 PU241 6.182E+03 6.257E+03 1.21 AM241 9.528E+00 9.797E+00 2.82 CM242 2.388E+03 2.263E+03

-5.25 CM244 2.602E+02 2.582E+02

-0.80 i

I i-

+-lk.*

TýQ Boldin is used if the last J-ositi-jr f

_r, j

1j Calc. No. L-003067, Rev. 1, Attachment A, Page A10 of A14

Input Deck islast.inp

-l

-i

-1 BAS RDA LIP LIB PHO RDA INP RDA INP RDA INP RDA TIT Grams of Heavy Metal per Fuel Batch PLACE FUEL into vectors -1, -2 and -3 0

0 0

0 1

2 3

657 658 659 9

3 0

1 42 0 0 0 10 READ FUEL COMPOSITION FOR BATCH 3

-1 1 1 1

1 READ FUEL COMPOSITION FOR BATCH 2

-2 1 1 1

1 READ FUEL COMPOSITION FOR BATCH 1

-3 1 1 1

1 IRRADIATION OF LaSalle Unit 1 CYCLE 1 FULL CORE MOV

-3 1

0 1.0 HED 1

CHARGE BATCH 1 FRESH RDA BUP IRP IRP IRP IRP IRP IRP IRP IRP IRP IRP IRP IRP IRP IRP BUP OPTL OPTA OPTF OUT MOV HED RDA BUP IRP IRP IRP IRP IRP IRP IRP IRP IRP IRP IRP IRP IRP IRP BUP BATCH 1 BURNUP IN CYCLE 1 50.0 100.0 150.0 200.0 250.0 300.0 350.0 400.0 450.0 500.0 550.0 600.0 650.0 711.0 4*8 5 4*8 5 4*8 5

-8 1

8 1

1010.5 1010.5 1010.5 1010.5 1010.5 1010.5 1010.5 1010.5 1010.5 1010.5 1010.5 1010.5 1010.5 1010.5 1

9 2

9 3

9 4

9 5

9 6

9 7

9 9

2 9

3 9

4 9

5 9

6 9

7 9

8 4

4 4

4 4

4 4

4 4

4 4

4 4

4 2

0 0

0 0

0 0

0 0

0 0

0 0

0 8 5 17*8 8 5 17*8 8 5 17*8

-1 0

0 1.0 BATCH 1 ONCE BURNED 1

CHARGE BATCH 1 BURNUP IN CYCLE 2 761.0 811.0 861.0 911.0 961.0 1011.0 1061.0 1111.0 1161.0 1211.0 1261.0 1311.0 1361.0 1422.0 952.2 952.2 952.2 952.2 952.2 952.2 952.2 952.2 952.2 952.2 952.2 952.2 952.2 952.2 1

9 2"

9 3

9 4

9 5

9 6

9 7

9 9

2 9

3 9

4 9

5 9

6 9

7 9

8 4

4 4

4 4

4 4

4 4

4 4

4 4

4 3

0 0

0 0

0 0

0 0

0 0

0 0

0 Calc. No. L-003067, Rev. 1, Attachment A, Page Al1 of A14

OUT MOV HED RDA BUP IRP IRP IRP IRP IRP IRP IRP IRP IRP IRP IRP IRP IRP IRP BUP

-OUT MOV MOV RDA MOV HED RDA BUP IRP IRP IRP IRP IRP IRP IRP IRP IRP IRP IRP IRP IRP IRP BUP OUT MOV HED

-8 1

-1 0 8

1 0

1.0 1

CHARGE BATCH 1 TWICE BURNED BATCH 1 BURNUP IN CYCLE 3 1472.0 1522.0 1572.0 1622.0 1672.0 1722.0 1772.0 1822.0 1872.0 1922.0 1972.0 2022.0 2072.0 2133.0

-8 1

2 -9 8 -10 408.6 408.6 408.6 408.6 408.6 408.6 408.6 408. 6 408.6 408.6 408.6 408.6 408.6 408.6

-1 0 0

1.0 0

1.0 1

9 2

9 3

9 4

9 5

9 6

9 7

9 9

2 9

3 9

4 9

5 9

6 9

7 9

8 4

4 4

4 4

4 4

4 4

4 4

4 4

4 3

0 0

0 0

0 0

0 0

0 0

0 0

0 1 100 EFPD PLACED IN TEMP VECTOR -9 1 EOC3 PLACED IN TEMP VECTOR -10 FRESH BATCH BATCH IN CYCLE 2 BATCH 2 BATCH 2 BURNUP

-2 1

0 1.0 1

CHARGE BATCH 2 BURNUP IN CYCLE 2 50.0 100.0 150.0 200.0 250.0 300.0 350.0 400.0 450.0 500.0 550.0 600.0 650.0 711.0 1589.2 1589.2 1589.2 1589.2 1589.2 1589.2 1589.2 1589.2 1589.2 1589.2 1589.2 1589.2 1589.2 1589.2 1

9 2

9 3

9 4

9 5

9 6

9 7

9 9

2 9

3 9

4 9

5 9

6 9

7 9

8 4

4 4

4 4

4 4

4 4

4 4

4 4

4 2

0 0

0 0

0 0

0 0

0 0

0 0

0

-8 1

-1 0 8

1 0

1.0 1

CHARGE BATCH 2 ONCE BURNED RDA BATCH 2 BURNUP IN CYCLE 3 BUP IRP IRP IRP IRP IRP IRP IRP IRP IRP IRP IRP 761.0 811.0 861.0 911.0 961.0 1011.0 1061.0 1111.0 1161.0 1211.0 1261.0 1487.8 1487.8 1487.8 1487.8 1487.8 1487.8 1487.8 1487.8 1487.8 1487.8 1487.8 1

9 2

9 3

9 4

9 5

9 6

9 2

9 3

9 4

9 5

9 6

9 4

4 4

4 4

4 4

4 4

4 4

3 0

0 0

0 0

0 0

0 0

0 Calc. No. L-003067, Rev. 1, Attachment A, Page A12 of A14

IRP IRP IRP BUP OUT ADD ADD MOV HED RDA BUP IRP IRP IRP IRP IRP IRP IRP IRP IRP IRP IRP IRP IRP IRP BUP OUT ADD ADD MOV MOV HED HED OUT END 1311.0 1361.0 1422.0 1487.8 1487.8 1487.8 9

7 9

7 9

8 4

0 4

0 4

0

-8 1 -1 0 2 -9 0

1.0 8 -10 0

1.0

-1 1

0 1.0 1

CHARGE BATCH BATCH BATCH 2

2 3

100 EFPD ADDED TO TEMP VECTOR -9 EOC3 ADDED TO TEMP VECTOR -10 FRESH BATCH 3 BURNUP IN CYCLE 3 50.0 100.0 150.0 200.0 250.0 300.0 350.0 400.0 450.0 500.0 550.0 600.0 650. 0 711.0 1589.2 1589.2 1589.2 1589.2 1589.2 1589.2 1589.2 1589.2 1589.2 1589.2 1589.2 1589.2 1589.2 1589.2 1

9 2

9 3

9 4

9 5

9 6

9 7

9 9

2 9

3 9

4 9

5 9

6 9

7 9

8 4

4 4

4 4

4 4

4 4

4 4

4 4

4 2

0 0

0 0

0 0

0 0

0 0

0 0

0

-8 2

8

-9

-10 2

-2 1

-1 0

-9 0

1.0

-10. 0 1.0 1

0 1.0 2

0 1.0 100 EFPD EOC 1

-1 0

CYCLE 3 @ 100 EFPD CYCLE 3 @ EOC 2

922350 1987371.94 922380 49334228.06 4

080000 6900124.43 0

0.0 0

2 922350 1987371.94 922380 49334228.06 4

080000 6900124.43 0

0.0 0

2 922350 1297312.24 922380 32204287.76 4

080000 4504247.89 0

0.0 0

END 0

0.0 0

0.0 0

0.0 U02 U02 U02 U02 U02 U02 Calc. No. L-003067, Rev. 1, Attachment A, Page A13 of A14

Job Batch File lsl ast.bat echo off echo *********************************************************************

echo ***********************************************************************

echo **

echo **0 R I G E N 2 echo **

Oak Ridge Isotope GENeration and Depletion Code echo **

Version 2.1 (8-1-91) echo **

echo ******************~**************************************~**************

echo **

echo Developed by:

Oak Ridge National Laboratory echo Chemical Technology Division echo **

echo Technical

Contact:

Scott B. Ludwig echo (615) 574-7916 FTS 624-7916 echo **

echo Distributed by:

Radiation Shielding Information Center (RSIC) echo Oak Ridge National Laboratory echo **

P.O. Box 2008 echo Oak Ridge, TN 37831 echo (615) 574-6176 FTS 624-6176 echo ***********************************************************************

echo *********************************************************************

pause echo ** Execution continuing...

echo ***********************************************************************

echo ***********************************-************************************

echo **

echo Version 2.1 (8-1-91) for mainframes and 80386 or 80486 PCs echo

  • copy isl ast.inp tape5.inp >nul REM (NOT USED IN THIS CASE) copy samp_2.u3 tape3.inp >nul copy

\\d-drive\\origen2l\\libs\\decay.lib+\\d-drive\\origen2l\\libs\\bwrue.lib tape9.inp >nul copy \\d-drive\\origen2l\\libs\\gxuo2brm.lib tapel0.inp >nul

\\d-drive\\origen2l\\code\\origen2 rem combine and save files from run copy tapel2.out+tape6.out lsl ast.u6 >nul copy tapel3.out+tapell.out lsl ast.ull

>nul ren tape7.out lsl ast.pch ren tapel5.out lsl ast.dbg ren tapel6.out lsl ast.vxs ren tape50.out isl ast.ech rem cleanup files del tape*.inp del tape*.out echo **********************************************************************

echo ******************* OR I G E N 2 - Version 2.1

  • echo *********************** Execution Completed ***************************

echo ***********************************************************************

echo on Caic. No. L-003067, Rev. 1, Attachment A, Page A14 of A14

LaSalle FHA 24 hr Decay - No CR or SGTS Filter Credit.o0 RADTRAD Version 3.03 (Spring 2001) run on 10/18/2007 at 15:36:48 File information Plant file

= P:\\Users\\Nuc\\Exelon EOC\\Discipline Files\\Process\\AST\\LaSalle AST\\LSCS FHA\\RADTRAD\\2007 RADTRAD\\LSCS FHA (Final) 24hr Delay -

No SGTS Filter Credit -

Purge CR Intake Flow -

GL Release -

No CREF Credit.psf Inventory file

= p:\\users\\nuc\\exelon eoc\\discipline files\\process\\ast\\lasalle ast\\lscs ast source terms\\lscs ast source terms for fha.nif Release file

= p:\\users\\nuc\\exelon eoc\\discipline files\\process\\ast\\lasalle ast\\liscs fha\\radtrad\\lasalle ast fha.rft Dose Conversion file

= c:\\program files\\radtrad3.03\\defaults\\fgrll&12.inp

  1. 4***

4* 4 4* #4*

4 4*

4 4*

4*

4 Radtrad 3.03 4/15/2001 LCS FHA - Diffuse Area to Limiting CR Intake with Purge Flow, EAB,

& LPZ 24 Hour Delay and No SGTS Filtration Credit -

CR Purge Intake Flow -

GL Release-No CREF Credit Nuclide Inventory File:

p:\\users\\nuc\\exelon eoc\\discipline files\\process\\ast\\lasalle ast\\lscs ast source terms\\lscs ast source terms for fha.nif Plant Power Level:

1. 5597E+01 Compartments:

3 Compartment 1:

Reactor Building 3

1. 0000E+00 0

0 0

0 0

Compartment 2:

Environment 2

0. 0000E+00 0

0 0

0 0

Compartment 3:

Control Room 1

1. 1750E+05 0

Calc. No.

L-003067, Rev.

1, Attachment B, Page B1 of B9

LaSalle FHA 24 hr Decay -

No CR or SGTS Filter Credit.oO 0

1 0

0 Pathways:

3 Pathway 1:

Reactor Building to Environment 1

2 2

Pathway 2:

Purge Flow Environment to 2

3 2

Pathway 3:

Purge Flow Control Room to 3

2 2

End of Plant Model File Scenario Description Name:

Control Room Environment Exhaust Plant Model Filename:

ACCEPT\\TEST1.PMF Source Term:

1 1

1.OOOOE+00 c:\\program files\\radtrad3.03\\defaults\\fgrll&12.inp p:\\users\\nuc\\exelon eoc\\discipline files\\process\\ast\\lasalle ast\\lscs fha\\radtrad\\lasalle ast fha.rft 2.4000E+01 1

0.0000E+00 9.7000E-01 3.OOOOE-02 Overlying Pool:

0 o.OOOOE+00 0

0 0

0 Compartments:

Compartment 1:

0 1

0 0

0 0

0 0

0 Compartment 2:

0 1

0 0

0 0

0 0

0 Compartment 3:

0 1.OOOOE+00 Calc. No. L-003067, Rev.

1, Attachment B, Page B2 of B9

LaSalle FHA 24 hr Decay -

No CR or SGTS Filter Credit.oO 0

0 0

0 1

1.4200E+04 4

2.4000E+01 0.0000E+00 2.4333E+01 0.0000E+00 2.8000E+01 0.OOOOE+00 4.8000E+01 0.OOOOE+00 0

0 Pathways:

3 Pathway 1:

0 0

0, 0

0 1

2 2.4000E+01 1.0000E-01 2.6000E+01 0.0000E+00 0

0 0

0 0

0 Pathway 2:

0 0

0 0

0 1

2 2.4000E+01 3.0000E+04 4.8000E+01 0.0000E+00 0

0 0

0 0

0 Pathway 3:

0 0

0 0

0 1

2 2.4000E+01 3.0000E+04 4.8000E+01 0.OOOOE+00 0

0 0

0 0

0 Dose Locations:

3 Location 1:

Exclusion Area Bndry o.OOOOE+00 o.OOOOE+/-00 o.OOOOE+00 o.OOOOE+/-00 o.OOOOE+00 o.OOOOE+0O o.OOOOE+00 o.OOOOE+00

1. OOOOE+02 o.OOOOE+00
0. OOOOE+00
0. OOOOE+00
0. OOOOE+00
0. 0000E+00
0. 0000E+00
0. 0000E+00
0. 0000E+00 o.OOOOE+00 1.OOOOE+02 o.OOOOE+00
0. OOOOE+00
0. OOOOE+00
0. OOOOE+00
0. 0000E+00
1. OOOOE+02
0. OOOOE+00 Calc.

No. L-003067, Rev.

1, Attachment B, Page B3 of B9

LaSalle FHA 24 hr Decay - No CR or SGTS Filter Credit.oO 2

1 3

2.4000E+01 5.4000E-04 2.4500E+01 5.4000E-04 4.8000E+01 O.OOOOE+00 1

2 2.4000E+01 3.5000E-04 4.8000E+01 O.OOOOE+00 0

Location 2:

Low Population Zone 2

1 3

2.4000E+01 2.2600E-05 2.4500E+01 2.2600E-05 4.8000E+01 0.OOOOE+00 1

2 2.4000E+01 3.5000E-04 4.8000E+01 0.OOOOE+00 0

Location 3:

Control Room 3

0 1

2 2.4000E+01 3.5000E-04 4.8000E+01 0.0000E+00 1

2 2.4000E+01 1.0000E+00 4.8000E+01 0.0000E+00 Effective Volume Location:

1 2

2.4000E+01 1.6700E-03 4.8000E+01 0.0000E+00 Simulation Parameters:

4 2.4000E+01 5.OOOOE-02 2.4500E+01 1.0000E-01 2.6000E+01 1.0000E+00 4.8000E+01 0.0000E+00 Output Filename:

P:\\Users\\Nuc\\Exelon EOC\\Discipline Files\\Process\\AST\\LaSalle AST\\LSCS FHA\\RADTRAD\\2007 RADTRAD\\LSCS FHA (Final) 24hr Delay -

No SGTS Filter Credit Purge CR Intake Flow -

GL Release -

No CREF Credit.ol 0

0 1

0 0

End of Scenario File Calc. No. L-003067, Rev.

1, Attachment B, Page B4 of B9

LaSalle FHA 24 hr Decay - No CR or SGTS Filter Credit.oO Calc. No.

L-003067, Rev.

1, Attachment B, Page B5 of B9

LaSalle FHA 24 hr Decay - No CR or SGTS Filter Credit.oO RADTRAD Version 3.03 (Spring 2001) run on 1/10/2008 at 17:18:26

        1. tf
  1. t f #####t ft
  1. t ft ft ft ft f#

ft ft ft ft #

ft ft f#

ft

        1. tf
        1. tf
  1. t
          1. ftf ft ft ft ft
          1. ftf ft
  1. t
  1. t ft ft
  1. t ft
  1. t
  1. t ft
  1. t ft ft ft
  1. t f#####f ft
  1. t ft ft ft ft ftfftfftf fttfttfttfttfftfftfftfftfftfftfftft tfttfttfttfttfttfttfftfftfftfftfftf ftftftfftfftfftfftf fttf Dose Output Exclusion Area Bndry Doses:

Time (h)

=

24.0000 Whole Body Delta dose (rem) 1.2715E-06 Accumulated dose (rem) 1.2715E-06 Low Population Zone Doses:

Time (h)

=

24.0000 Delta dose (rem)

Accumulated dose (rem)

Control Room Doses:

Time (h)

=

24.0000 Delta dose (rem).

Accumulated dose (rem)

Whole Body 5.3213E-08 5.3213E-08 Whole Body 1.3279E-12 1.3279E-12 Exclusion Area Bndry Doses:

Time (h) =

24.3330 Whole Body Delta dose (rem) 3.6394E-01 Accumulated dose (rem) 3..6394E-01 Thyroid 1.0562E-04 1.0562E-04 Thyroid 4.4206E-06 4.4206E-06 Thyroid 2.5020E-09 2.5020E-09 Thyroid 3.0412E+01 3.0412E+01 Thyroid 1.2728E+00 1.2728E+00 Thyroid

8. 2777E+01 8.2777E+01 Thyroid
3. 0125E+00
3. 3424E+01 TEDE 4.5235E-06 4.5235E-06 TEDE 1.8932E-07 1.8932E-07 TEDE
7. 8361E-II
7. 8361E-l1 TEDE 1.3002E+00 1.3002E+00 TEDE
5. 4414E-02 5.4414E-02 TEDE 2. 5917E+00
2. 5917E+00 TEDE 1.2809E-01 1.4282E+00 Low Population Zone Doses:

Time (h)

=

24.3330 Delta dose (rem)

Accumulated dose (rem)

Control Room Doses:

Time (h)

=

24.3330 Delta dose (rem)

Accumulated dose (rem)

Whole Body 1.5232E-02 1.5232E-02 Whole Body 4.3555E-02 4.3555E-02 Exclusion Area Bndry Doses:

Time (h)

=

24.5000 Delta dose (rem)

Accumulated dose (rem)

Whole Body 3.5388E-02 3.9933E-01 Low Population Zone Doses:

Time (h)

=

24.5000 Whole Body Thyroid TEDE Calc. No. L-003067, Rev. 1, Attachment B, Page B6 of B9

LaSalle FHA 24 hr Decay - No CR or SGTS Filter Credit.oo Delta dose (rem)

Accumulated dose (rem)

Control Room Doses:

Time (h)

=

24.5000 Delta dose (rem)

Accumulated dose (rem) 1.4811E-03 1.6713E-02 Whole Body 7.7723E-03 5.1327E-02 Exclusion Area Bndry Doses:

Time (h)

=

26.0000 Delta dose (rem)

Accumulated dose (rem)

Whole Body 2.0187E-02 4.1952E-01 Low Population Zone Doses:

Time (h)

=

26.0000 Delta dose (rem)

Accumulated dose (rem)

Control Room Doses:

Time (h)

=

26.0000 Delta dose (rem)

Accumulated dose (rem)

Whole Body 8.4486E-04 1.7558E-02 Whole Body 4.6397E-03 5.5967E-02 Exclusion Area Bndry Doses:

Time (h)

=

28.0000 Delta dose (rem)

Accumulated dose (rem)

Whole Body 0.OOOOE+00 4.1952E-01 1.2608E-01 1.3989E+00 Thyroid 1.5006E+01

9. 7784E+01 Thyroid 1.7442E+00
3. 5169E+01 Thyroid 7.2999E-02 1.4719E+00 Thyroid 9.0924E+00 1.0688E+02 Thyroid 0.0000E+00 3. 5169E+01 Thyroid 0.0000E+00
1. 4719E+00 Thyroid 5.0408E-04 1.0688E+02 Thyroid 0.0000E+00 3.5169E+01 Thyroid 0.0000E+00 1.4719E+00 Thyroid 2.4566E-17 1.0688E+02 5.3608E-03 5.9775E-02 TEDE
4. 6955E-01
3. 0612E+00 TEDE 7.3845E-02 1.5021E+00 TEDE 3.0905E-03 6.2865E-02 TEDE 2.8435E-01 3.3456E+00 TEDE 0.OOOOE+00 1.5021E+00 TEDE 0.OOOOE+00 6.2865E-02 TEDE 1.5721E-05 3.3456E+00 TEDE 0.OOOOE+00 1.5021E+00 TEDE 0.OOOOE+00 6.2865E-02 TEDE
7. 6396E-19 3.3456E+00 Low Population Zone Doses:

Time (h)

=

28.0000 Delta dose (rem)

Accumulated dose (rem)

Control Room Doses:

Time (h)

=

28.0000 Delta dose (rem)

Accumulated dose (rem)

Whole Body 0.OOOOE+00 1.7558E-02 Whole Body 2.3681E-07 5.5967E-02 Exclusion Area Bndry Doses:

Time (h)

=

48.0000 Delta dose (rem)

Accumulated dose (rem)

Whole Body 0.OOOOE+00 4.1952E-01 Low Population Zone Doses:

Time (h) =

48.0000 Delta dose (rem)

Accumulated dose (rem)

Control Room Doses:

Time (h)

=

48.0000 Delta dose (rem)

Accumulated dose (rem)

Whole Body 0.OOOOE+00 1.7558E-02 Whole Body 1.0398E-20 5.5967E-02 147 1-131 Summary Calc.

No. L-003067, Rev.

1, Attachment B, Page B7 of B9

LaSalle FHA 24 hr Decay -

No CR or SGTS Filter Credit.oO Time (hr) 24.000 24.300 24.333 24.500 24.800 25.100 25.400 25.700 26.000 26.300 26.600 26.900 27.200 27.500 27.800 28.000 28.300 28.600 28.900 29.200 29.500 29.800 30.100 30.400 30.700 31.000 31.300 31.600 31.900 32.200 32.500 32.800 33.100 33.400 33.700 34.000 34.300 48.000 Reactor Building 1-131 (Curies) 1.5499E+02 2.5593E+01 2.0993E+01 7.7029E+00 1.2719E+00 2.1002E-01 3.4679E-02 5.7262E-03 9.4551E-04 9.4449E-04 9.4347E-04 9.4246E-04 9.4144E-04 9.4043E-04 9.3942E-04 9.3874E-04 9.3773E-04 9.3672E-04 9.3571E-04 9.3470E-04 9.3370E-04 9.3269E-04 9.3169E-04 9.3068E-04 9.2968E-04 9.2868E-04 9.2768E-04 9.2668E-04 9.2568E-04 9.2468E-04 9.2369E-04 9.2269E-04 9.2170E-04 9.2071E-04 9.1971E-04 9.1872E-04 9.1773E-04 8.7366E-04 Environment 1-131 (Curies)

4. 6498E-04
1. 2931E+02
1. 3391E+02
1. 4719E+02
1. 5362E+02
1. 5468E+02
1. 5485E+02 1.5488E+02 1.5489E+02 1.5489E+02 1.5489E+02 1.5489E+02 1.5489E+02 1.5489E+02 1.5489E+02 1.5489E+02 1.5489E+02 1.5489E+02 1.5489E+02 1.5489E+02 1.5489E+02 1.5489E+02 1.5489E+02 1.5489E+02 1.5489E+02 1.5489E+02 1.5489E+02 1.5489E+02 1.5489E+02 1.5489E+02 1.5489E+02 1.5489E+02 1.5489E+02 1.5489E+02 1.5489E+02 1.5489E+02 1.5489E+02 1.5489E+02 Control Room 1-131 (Curies)
1. 0994E-05
3. 6581E-01
3. 0523E-01
1. 1615E-01
1. 9351E-02
3. 1971E-03
5. 2792E-04
8. 7171E-05 1.4394E-05
1. 4514E-07 1.4635E-09 1.4757E-11 1.4880E-13
1. 5005E-15
1. 5130E-17
7. 0617E-19
7. 1206E-21 7.1801E-23
7. 2400E-25
7. 3004E-27
7. 3614E-29 7.4228E-31 7.4848E-33 7.5473E-35 7.6103E-37 7.6738E-39 7.7378E-41 7.8024E-43 7.8676E-45' 7.9332E-47 7.9995E-49 8.0662E-51 8.1336E-53 8.2015E-55 8.2699E-57 8.3390E-59 8.4086E-61 5.6711-152 Cumulative Dose Summary Exclusion Area Bndry Low Population Zone Time (hr) 24.000 24.300 24.333 24.500 24.800 25.100 25.400 25.700 26.000 26.300 26.600 26.900 27.200 27.500 27.800 28.000 28.300 Thyroid (rem) 0.OOOOE+00
2. 9369E+01
3. 0412E+01 3.3424E+01 3.4881E+01 3.5121E+01 3.5161E+01 3.5167E+01
3. 5169E+01
3. 5169E+01 3.5169E+01 3.5169E+01
3. 5169E+01
3. 5169E+01
3. 5169E+01 3.5169E+01
3. 5169E+01 TEDE (rem) 0.0000E+00 1.2557E+00 1.3002E+00 1.4282E+00 1.4900E+00
1. 5001E+00
1. 5018E+00 1.5020E+00 1.5021E+00 1.5021E+00 1.5021E+00 1.5021E+00 1.5021E+00 1.5021E+00 1.5021E+00 1.5021E+00 1.5021E+00 Thyroid (rem) 0.OOOOE+00 1.2291E+00 1.2728E+00 1.3989E+00 1.4598E+00 1.4699E+00 1.4716E+00 1.4718E+00 1.4719E+00 1.4719E+00 1.4719E+00 1.4719E+00 1.4719E+00 1.4719E+00 1.4719E+00
1. 4719E+00 1.4719E+00 TEDE (rem) 0.OOOOE+00 5.2555E-02 5.4414E-02 5.9775E-02 6.2359E-02 6.2783E-02 6.2852E-02 6.2864E-02 6.2865E-02 6.2865E-02 6.2865E-02 6.2865E-02 6.2865E-02 6.2865E-02 6.2865E-02 6.2865E-02 6.2865E-02 Control Thyroid (rem) 0.OOOOE+00
7. 7744E+01
8. 2777E+01 9.7784E+01
1. 0537E+02 1.0663E+02
1. 0684E+02
1. 0687E+02
1. 0688E+02 1.0688E+02 1.0688E+02
1. 0688E+02 1.0688E+02 1.0688E+02
1. 0688E+02
1. 0688E+02 1.0688E+02 Room TEDE (rem) 0.OOOOE+00 2.4342E+00 2.5917E+00 3.0612E+00 3.2986E+00 3.3379E+00 3.3443E+00 3.3454E+00 3.3456E+00 3.3456E+00 3.3456E+00 3.3456E+00 3.3456E+00 3.3456E+00 3.3456E+00 3.3456E+00 3.3456E+00 Calc. No.

L-003067, Rev.

1, Attachment B, Page B8 of B9

LaSalle FHA 24 hr Decay -

No CR or SGTS Filter Credit.oO 28.600 28.900 29.200 29.500 29.800 30.100 30.400 30.700 31.000 31.300 31.600 31.900 32.200 32.500 32.800 33.100 33.400 33.700 34.000 34.300 48.000

3. 5169E+01 3.5169E+01
3. 5169E+01
3. 5169E+01
3. 5169E+01
3. 5169E+01
3. 5169E+01
3. 5169E+01
3. 5169E+01
3. 5169E+01 3.5169E+01
3. 5169E+01
3. 5169E+01 3. 5169E+01 3.5169E+01
3. 5169E+01 3. 5169E+01
3. 5169E+01 3.5169E+01
3. 5169E+01
3. 5169E+01
1. 5021E+00
1. 5021E+00
1. 5021E+00
1. 5021E+00
1. 5021E+00 1.5021E+00
1. 5021E+00
1. 5021E+00 1.5021E+00 1.5021E+00
1. 5021E+00 1.5021E+00 1.5021E+00 1.5021E+00 1.5021E+00 1.5021E+00 1.5021E+00 1.5021E+00
1. 5021E+00 1.5021E+00
1. 5021E+00 1.4719E+00 1.4719E+00 1.4719E+00 1.4719E+00 1.4719E+00 1.4719E+00 1.4719E+00 1.4719E+00 1.4719E+00 1.4719E+00 1.4719E+00 1.4719E+00 1.4719E+00 1.4719E+00 1.4719E+00 1.4719E+00 1.4719E+00 1.4719E+00 1.4719E+00 1.4719E+00
1. 4719E+00 6.2865E-02 6.2865E-02 6.2865E-02 6.2865E-02 6.2865E-02 6.2865E-02 6.2865E-02 6.2865E-02 6.2865E-02 6.2865E-02 6.2865E-02 6.2865E-02 6.2865E-02 6.2865E-02 6.2865E-02 6.2865E-02 6.2865E-02 6.2865E-02 6.2865E-02 6.2865E-02 6.2865E-02 1.0688E+02 1.0688E+02 1.0688E+02 1.0688E+02 1.0688E+02 1.0688E+02 1.0688E+02 1.0688E+02 1.0688E+02 1.0688E+02 1.0688E+02 1.0688E+02 1.0688E+02 1.0688E+02 1.0688E+02 1.0688E+02 1.0688E+02 1 0688E+02 1.0688E+02 1.0688E+02 1.0688E+02 3.3456E+00 3.3456E+00 3.3456E+00 3.3456E+00 3.3456E+00 3.3456E+00 3.3456E+00 3.3456E+00 3.3456E+00 3.3456E+00 3.3456E+00 3.3456E+00 3.3456E+00 3.3456E+00 3.3456E+00 3.3456E+00 3.3456E+00 3.3456E+00 3.3456E+00 3.3456E+00 3.3456E+00 Worst Two-Hour Doses Exclusion Time (hr) 24.0 Area Bndry Whole Body (rem) 4.1952E-01 Thyroid (rem) 3.5169E+01 TEDE (rem)
1. 5021E+00 Calc. No.

L-003067, Rev.

1, Attachment B, Page B9 of 39

LaSalle Generating Station AST Source Terms for FHA.nif Nuclide Inventory Name:

LaSalle County Station Power Level:

0.1000E+01 Nuclides:

60 Nuclide 001:

Co-58 7

0.6117120000E+07 0.5800E+02 0.1529E+03 none 0.OOOOE+00 none 0.OOOOE+00 none 0.OOOOE+00 Nuclide 002:

Co-60 7

0.1663401096E+09 0.6000E+02 0.1830E+03 none 0.0000E+00 none 0.OOOOE+00 none 0.0000E+00 Nuclide 003:

Kr-85 1

0.3382974720E+09 0.8500E+02 0.7892E+03 none 0.0000E+00 none 0.OOOOE+00 none 0.0000E+00 Nuclide 004:

Kr-85m 1

Source Terms per this calculation (LSCS)

AST -

in Ci/MW 2.0*LOCA Value for FHA 0.1612800000E+05 0.8500E+02 0.8313E+04 Kr-85 0.2100E+ 00 none 0.0 none 0.0 Nuclide 005:

Kr-87 OOOE+00 OOOE+00 1

0.4578000000E+04 0.8700E+02 0.1633E+05 Rb-87 0.1000E+01 none 0.OOOOE+00 none 0.OOOOE+00 Nuclide 006:

Kr-88 1

0.1022400000E+05 0.8800E+02 0.2303E+05 Rb-88 0.1000E+01 none 0.OOOOE+00 Calc. No. L-003067, Rev. 1, Attachment C, Page C1 of C10

LaSalle Generating Station AST Source Terms for FHA.nif none 0.OOOOE+00 Nuclide 007:

Rb-86 3

0.1612224000E+07 0.8600E+02 0.6518E+02 none 0.0000E+00 none 0.OOOOE+00 none 0.OOOOE+00 Nuclide 008:

Sr-89 5

0.4363200000E+07 0.8900E+02 0.2798E+05 none 0.0000E+00 none 0.0000E+00 none 0.0000E+00 Nuclide 009:

Sr-90 5

0.9189573120E+09 0.9000E+02 0.3178E+04 Y-90 0.1000E+01 none 0.0000E+00 none 0.0000E+00 Nuclide 010:

Sr-91 5

0.3420000000E+05 0.9100E+02 0.3801E+05 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 0.4017E+05 Y-92 0.1000E+01 none 0.OOOOE+00 none 0.OOOOE+00 Nuclide 012:

Y-90 9

0.2304000000E+06 0.9000E+02 0.3272E+04 none 0.OOOOE+00 none 0.OOOOE+00 none 0.OOOOE+00 Nuclide 013:

Y-91 9

0.5055264000E+07 Calc. No. L-003067, Rev. 1, Attachment C, Page C2 of C10

LaSalle Generating Station AST Source Terms for FHA.nif

0. 9100E+02 0.3448E+05 none O.OOOOE+00 none 0.OOOOE+00 none 0.OOOOE+00 Nuclide 014:

Y-92 9.

0.1274400000E+05 0.9200E+02 0.4029E+05 none 0.0000E+00 none 0.OOOOE+00 none 0.OOOOE+00 Nuclide 015:

Y-93 9

0.3636000000E+05 0.9300E+02 0.4526E+05 Zr-93 0.1000E+01 none 0.0000E+00 none 0.0000E+00 Nuclide 016:

Zr-95 9

0.5527872000E+07 0.9500E+02 0.4489E+05 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 0.4657E+05 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 0.4512E+05 none 0.OOOOE+00 none 0.OOOOE+00 none 0.OOOOE+00 Nuclide 019:

Mo-99 7

0.2376000000E+06 0.9900E+02 0.5078+05 Tc-99m 0.8800E+00 Tc-99 0.1200E+00 none 0.OOOOE+00 Caic. No. L-003067, Rev. 1, Attachment C, Page C3 of CIO

LaSalle Generating Station AST Source Terms for FHA.nif Nuclide 020:

Tc-99m 7

0.2167200000E+05 0.9900E+02 0.4447E+05 Tc-99 0.1000E+01 none 0.0000E+00 none 0.OOOOE+00 Nuclide 021:

Ru-103 7

0.3393792000E+07 0.1030E+03 0.4242E+05 Rh-103m 0.1000E+01 none 0.0000E+00 none 0.0000E+00 Nuclide 022:

Ru-105 7

0.1598400000E+05 0.1050E+03 0.2966E+05 Rh-105 0.1000E+01 none 0.0000E+00 none 0.0000E+00 Nuclide 023:

Ru-106 7

0.3181248000E+08

0. 1060E+03 0.1760E+05 Rh-106 0.1000E+01 none 0.0000E+00 none 0.0000E+00 Nuclide 024:

Rh-105 7

0.1272960000E+06 0.1050E+03 0.2752E+05 none 0.0000E+00 none 0.0000E+00 none 0.0000E+00 Nuclide 025:

Sb-127 4

0.3326400000E+06 0.1270E+03 0.2944E+04 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 CaIc. No. L-003067, Rev. 1, Attachment C, Page C4 of CIO

LaSalle Generating Station AST Source Terms for FHA.nif 0.8725E+04 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 0.2918E+04 none 0.0000E+00 none 0.0000E+00 none 0.0000E+00 Nuclide 028:

Te-127m 4

0.9417600000E+07 0.1270E+03 0.3909E+03 Te-127 0.9800E+00 none 0.OOOOE+00 none 0.0000E+00 Nuclide 029:

Te-129 4

0.4176000000E+04 0.1290E+03 0.8584E+04 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. 1279E+04 Te-129 0.6500E+00 1-129 0.3500E+00 none 0.0000E+00 Nuclide 031:

Te-131m 4

0.1080000000E+06 0.1310E+03 0.3892E+04 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 0.3829E+05 1-132 0.1000E+01 none 0.OOOOE+00 none 0.OOOOE+00 Nuclide 033:

Calc. No. L-003067, Rev. 1, Attachment C, Page C5 of C10

LaSalle Generating Station AST Source Terms for FHA.nif 1-131 2

0.6946560000E+06 0.1310E+03 0.4312E+05 1.6*LOCA Value for FHA Xe-131m 0.1100E-01 none 0.0000E+00 none 0.O000E+00 Nuclide 034:

1-132 2

0.8280000000E+04 0.1320E+03 0.3889E+05 none 0.0000E+00 none 0.0000E+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 0.0000E+00 Nuclide 036:

1-134 2

0.3156000000E+04 0.1340E+03 0.6165E+05 none 0.0000E+00 none 0.0000E+00 none 0.0000E+00 Nuclide 037:

1-135 2

0.2379600000E+05 0.1350E+03 0.5192E+05 Xe-135m. 0.1500E+00 Xe-135 0.8500E+00 none 0.OOOOE+00 Nuclide 038:

Xe-133 1

0.4531680000E+06 0.1330E+03 0.5491E+05 none 0.0000E+00 none 0.OOOOE+00 none 0.OOOOE+00 Nuclide 039:

Xe-135 1

0.3272400000E+05 0.1350E+03 0.2228E+05 CaIc. No. L-003067, Rev. 1, Attachment C, Page C6 of CIO

LaSalle Generating Station AST Source Terms for FHA.nif Cs-135 0.1000E+01 none 0.OOOOE+00 none 0.OOOOE+00 Nuclide 040:

Cs-134 3

0.6507177120E+08 0.1340E+03 0.7280E+04 none 0.0000E+00 none 0.OOOOE+00 none 0.OOOOE+00 Nuclide 041:.

Cs-136 3

0.1131840000E+07 0.1360E+03 0.2027E+04 none 0.0000E+00 none 0.OOOOE+00 none 0.0000E+00 Nuclide 042:

Cs-137 3

0.9467280000E+09 0.1370E+03 0.4538E+04 Ba-137m 0.9500E+00 none 0.OOOOE+00 none 0.0000E+00 Nuclide 043:

Ba-139 6

0.4962000000E+04 0.1390E+03 0.5084E+05 none 0.0000E+00 none 0.0000E+00 none 0.OOOOE+00 Nuclide 044:

Ba-140 6

0.1100736000E+07 0.1400E+03 0.4896E+05 La-140 0.1000E+01 none 0.OOOOE+00 none 0.OOOOE+00 Nuclide 045:

La-140 9

0.1449792000E+06 0.1400E+03 0.5019E+05 none 0.OOOOE+00 none 0.OOOOE+00 none 0.OOOOE+00 Nuclide 046:

La-141 CaIc. No. L-003067, Rev. 1, Attachment C, Page C7 of CIO

LaSalle Generating Station AST Source Terms for FHA.nif 9

0.1414800000E+05 0.1410E+03 0.4640E+05 Ce-141 0.1000E+01 none 0.0000E+00 none 0.0000E+00 Nuclide 047:

La-142 9

0.5550000000E+04 0.1420E+03 0.4532E+05 none 0.0000E+00 none 0.0000E+00 none 0.0000E+00 Nuclide 048:

Ce-141 8

0.2808086400E+07 0.1410E+03 0.4492E+05 none 0.OOOOE+00 none 0.OOOOE+00 none 0.OOOOE+00 Nuclide 049:

Ce-143 8

0.1188000000E+06 0.1430E+03 0.4427E+05 Pr-143 0.1000E+01 none 0.0000E+00 none 0.0000E+00 Nuclide 050:

Ce-144 8

0.2456352000E+08 0.1440E+03 0.3596E+05 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 0.4293E+05 none 0.OOOOE+00 none 0.OOOOE+00 none 0.OOOOE+00 Nuclide 052:

Nd-147 9

0.9486720000E+06 0.1470E+03 0.1838E+05 Pm-147 0.1000E+01 Calc. No. L-003067, Rev. 1, Attachment C, Page C8 of C10

LaSalle Generating Station AST Source Terms for FHA.nif none 0.OOOOE+00 none 0.OOOOE+00 Nuclide 053:

Np-239 8

0.2034720000E+06 0.2390E+03 0.5540E+06 Pu-239 0.1000E+01 none 0.OOOOE+00 none 0.0000E+00 Nuclide 054:

Pu-238 8

0.2768863824E+10 0.2380E+03 0.1796E+03 U-234 0.1000E+01 none 0.OOOOE+00 none 0.OOOOE+00 Nuclide 055:

Pu-239 8

0.7594336440E+12 0.2390E+03 0.1207E+02 U-235 0.1000E+01 none 0.OOOOE+00 none 0.OOOOE+00 Nuclide 056:

Pu-240 8

0.2062920312E+12 0.2400E+03 0.1308E+02 U-236 0.1000E+01 none 0.OOOOE+00 none 0.OOOOE+00 Nuclide 057:

Pu-241 8

0.4544294400E+09 0.2410E+03 0.6257E+04 U-237 0.2400E-04 Am-241 0.1000E+01 none 0.OOOOE+00 Nuclide 058:

Am-241 9

0.1363919472E+II 0.2410E+03 0.9797E+01 Np-237 0.1000E+01 none 0.OOOOE+00 none 0.OOOOE+00 Nuclide 059:

Cm-242 9

Calc. No. L-003067, Rev. 1, Attachment C, Page C9 of C10

LaSalle Generating Station AST Source Terms for FHA.nif 0.1406592000E+08 0.2420E+03 0.2388E+04 Pu-238 0.1000E+01 none O.OOOOE+00 none 0.OOOOE+00 Nuclide 060:

Cm-244 9

0.5715081360E+09 0.2440E+03 o0.2602E+03 Pu-240 0.1000E+01 none Q.0000E+00 none 0.OOOE+00 End of Nuclear Inventory File Calc. No. L-003067, Rev. 1, Attachment C, Page C10 of C10

LaSalle Generating Station AST FHA.rft Release Fraction and Timing Name:

LaSalle Generating Station FHA, 10xlO bundle Nobles=I=5% & pool I DF=200, Cs DF=infinity Duration (h):

0. 10OE-05 Noble Gases:

5.000E-02 Iodine:

2. 5000E-04 Cesium:

o 0OOOE+00 Tellurium:

o

.OOOE+00 Strontium:

o.OOOOE+00 Barium:

o.000E+00 Ruthenium:

o.OOOE+00 Cerium:

o.000E+00 Lanthanum:

o.OOOE+00 Non-Radioacti 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 0.OOO0E+00 0.0OOOE+00 0.OO0E+00 0.0000E+00 0.0000E+00 0.0000E+00 0.OOOOE+00 0.0000E+00 0.OOOOE+00 0.0000E+00 0.OOOOE+00 0.OOOOE+00 0.0000E+00 0.0000E+00 0.0000E+00 ve Aerosols (kg):

0.0000E+00 o.OOOOE+00 o.OOOOE+00

o. 0000E+00
o.

0000E+00 0 0000E+00 0 0000E+00 0 0000E+00 0 0000E+00 0 OOOOE+00 0.0000E+00 0.0000E+00 End of Release File 0.OOOOE+00 0.OOOOE+00 Calc. No. L-003067, Rev. 1, AttachmentD, Page DI of DI

A 8

C D

E F

G H

K L

M

]

N 0

1 LSCS Fuel Handling Accident Assessment of Limiting Event 2

This

Attachment:

3

[a] Evaluates water coverage for FHAs over the Reactor Well and over the Spent 4

Fuel Pool.

5

[b] Evaluates impact of water coverages of less than 23 feet for purposes of pool 6

DF determination.

7 1

1_

8 Baseline R.G. 1.183 based Analysis of DFs 9

RG 1.183 RG 1.183 10 Water RG 1.183 RG 1.183 Inorganic Organic 11 Coverage Inorganic Organic Iodine Iodine DF Overall 1

12 (feet)

Iodine DF Iodine DF Fraction Fraction DF 13 23 500 1

0.9985 0.0015 286.0 Case 1: Inorganic Iodine DF Guidance Controlling 14 23 285.3 1

0.9985 0.0015 200.0 Case 2: Overall DF Guidance Controlling 15,

16 17 DFs determined per Burley Paper with R.G. 1.183 Case 1 assumptions 18 RG 1.183 RG 1.183 t

19 Water RG 1.183 RG 1.183 Inorganic Organic I

20 Coverage Inorganic Organic Iodine Iodine DF Overall 21 (feet)

Iodine DF Iodine DF Fraction Fraction DF 22 23 500 1

0.9985 0.0015 286.0 capped at 200 23 22.5 436.8 1

0.9985 0.0015 264.1 capped at 200 24 22 381.6 1

0.9985 0.0015 242.9 capped at 200 1

25 21.5 333.4 1

0.9985 0.0015 222.5 capped at 200 26 21 291.3 1

0.9985 0.0015 202.9 capped at 200 27 20.5 254.5 1

0.9985 0.0015 184.4 I

28 20 222.3 1

0.9985 0.0015 166.9 1

29 19.5 194.2 1

0.9985 0.0015 150.6 30 19 169.7 1

0.9985 0.0015 135.4 31 All water coverages are morethan 21 feet. Therefore, the 200 DF is conservative for all cases.

32 I

I 33 DFs determined per Burley Paper with R.G. 1.183 Case 2 assumptions 34 RG1.183 RG1.183 35 Water RG 1.183 RG 1.183 Inorganic Organic 36 Coverage Inorganic Organic Iodine Iodine DF Overall 37 (feet)

Iodine OF Iodine DF Fraction Fraction DF 38 23 285.3 1

0.9985 0.0015 200.0 39 22.5 252.3 1

0.9985 0.0015 183.2 40 22 223.1 1

0.9985 0.0015 167.4 41 21.5 197.3 1

0.9985 0.0015 152.4 42 21 174.5 1

0.9985 0.0015 138.5 43 20.5 154.3 1

0.9985 0.0015 125.5 44 20 136.5 1

0.9985 0.0015 113.4 45 19.5 120.7 1

0.9985 0.0015 102.3 46 19 106.7 1

0.9985 0.0015 92.1 47 20.868 168.9 1

0.9985 0.0015 134.9 48 21.333 189.4 1

0.9985 0.0015 147.7 49 117.000 Total number of failed rods (Full or Part-length), from pg E-5 50 92.000 Failed Rods (Full or Part-length) in dropped bundle, from page E-5 51 25.000 Failed Rods (Full or Part-length) in impacted bundle, from page E-5 52 0.949 Equivalency fraction of full-length rods in GE14 10x10 bundle 53 87.330 Failed Equivalent full-length rods in dropped bundle, with DF from Row 47 applicable 54 23.731 Failed Equivalent full-length rods in impacted bundle, with DF from Row 48 applicable 55 1

I 56 Overall DF weighted by in rack (impacted) vs. dropped bundles:

137.7 57 For Case 2, overall % of 200 DF 68.8%

59 Fuel failure over SFP vs. over the reactor well.

68.0%

60 based on Page E-5 I

I I

61 Therefore, the drop over the reactor well is bounding when a DF of 200 is used.

Calc. No. L-003067, Rev. I Attachment E, Page E1 of E5

A B

C D

E F

G H

K 1 LSCS Fuel Handlin Accident 2 This

Attachment:

3

[a] Evaluates water coverage for FHAs over the Reactor Well and over the Spent Fuel Pool.

4

[b] Evaluates impact of water coverages of less than 23 feet for purposes of pool DF determination.

5

[c] Justifies that a FHA over the Reactor Well is the limiting event.

6 8 Baseline R.G. 1.183 based Analysis o_

9 RG 1.183 RG 1.183 10 Water RG 1.183 RG 1.183 Inorganic Organic 11 Coverage Inorganic Organic Iodine Iodine DF Overall 12 (feet)

Iodine DF Iodine DF Fraction Fraction DF 13 23 000 1

0.9985 0.0015

=1/(D13/B13+E13/C13)

Case 1: Inorganic lo 14 23 1285.3 1

0.9985 0.0015

=1,;L 14!814÷E14iC14)

Case 2: Overall DF 15 1

16 1

17 DFs determined per Burley Paper witli 18 __RG 1.183 RG 1.183 19 Water I

RG 1.183 RG 1.183 Inorganic Organic 20 Coverage Inorganic Organic Iodine Iodine DF Overall 21 (feet)

=oieD Iodine DF Fraction Fraction DF 22 23 1500 10.9985 0.0015

=1/(D22/B22+E22/C22) capped at 200 23 =A22-0.5

=B$22^ A23/A$22) 1 0.9985 0.0015

=1/(D23/B23+E23/C23) capped at 200 24 =A23-0.5

=B$22^(A24/A$22) 1 0.9985 0.0015

=lIca241B24+E241C24 copped at 200 25 =A24-0.5

=B$22^ A25/A$22) 1 0.9985 0.0015

=11ca25/B25+E25/C25 copped at 200 26 =A25-0.5

=B$22^(A26/A$22) 1 0.9985 0.0015

=1/(D26/B26+E26/C26) capped at 200 27 =A26-0.5

=B$22^(A27/A$22) 1 0.9985 0.0015

=11(D271B27+E271C27) 28 =A27-0.5

=B$22^{A28/A$22) 1 0.9985 0.0015

=1/(D28/B28+E28/C28 29 =A28-0.5

=B$22^(A29/A$22) 1 0.9985 0.0015

=1/(D29/B29+E29/C29) 30 =A29-0.5

=B$22^(A30/A$22) 1 0.9985 0.0015

=1/(D30/B30+E30/C30) 31 All water coverages are more than 21 fei 321 33 DFs determined per Burley Paper witH 34 RG 1.183 RG 1.183 35 Water RG 1.183 RG 1.183 Inorganic Organic 36 CoveraInorganic Organic Iodine Iodine DF Overall 37Iodine F

Iodine DF Fraction Fraction DF 38 23 283 0.9985 0.0015

=1/1D38/B38+E38/C38) 39 =A38-0.5

=B$38^(A39/A$38) 1 0.9985 0.0015

=1/(D39/B39+E39/C39) 40 =A39-0.5

=B$38^{A40/A$38) 1 0.9985 0.0015

=1/1D40/B40+E40/C40) 41 =A40-0.5 i=B$38loA41/A$38) o 0.9985 0.0015

=11 (D41_/1341

+E411

/C41) 42 =A41-0.5

=B

^A42nA$3 0.9985 f0.0015r 43 =A42-0.5 1=B$ý38^A43/A$38) 0.9985 0.0015

=1D3B3E3C3 44 =A43-0.5

=B$38^(A44/A$38) 10.9985 0.0015

=-11(D441B44+E441C44) 45 =A44-0.5

=B$38^(A45/A$38) 10.9985 0.0015

=1/(D45/B45+E45/C45) 46 =A45-0.5

=B$38^(A46/A$38) 10.9985 0.0015

=1D6B6E6C6 47 20.868

=B$38^(A47/A$38) 1 0.9985 o,.0015

=1/(D471B47+E471C47) 48 21.333

=B$38^(A48/A$38) 10.9985 b.0015

=11(D481B48+E481C48) 49 117 Total number of failed ro 50 92 Failed Rods (Full or Part-51 =A49-A50 Failed Rods (Full or Part.

52 =87.33/92 Equivalency fraction of fu 53 =A52*A50 Failed Equivalent full-len 54 =A51 *A52 Failed Equivalent full-len 55 56 by in rack (impacted)

=((A53*F47)+(A54*Ft 571 2, overall % of 200 DF =K56/200 58 59 60 61 Calc. No. L-003067, Rev. I Attachment E, Page E2 of E5

AI B

I C

I D

F G

H 1 LSCS Fuel Handling Accident Assessment of Limiting Event 3

Primar Data Derived Values 4 Elevations MSL (ft)

Reference Spreadsheet #

Description Formula Value Value MSL (ft) 5 A Top of Grapple Rail 843.583 LaSalle Drawing S-757, Sht 8, Rev 18 Top of Bail to Top of Rod (in)

= P - N 8.780 96 Refuel Floor 843.500 LaSalle Drawing S-784, Sht 2, Rev.

19A Ul Top of Bail in SFP

= D + 0 + R + J 820.081 7 C Bottom of Cattle Chute 819.729 LaSalle Drawing S-784, Sht 2, Rev.

19B U2 Top of Bail in SFP

= 0+ T + U + J 820.290 0

[Bottom of SFP 804.750 S-784 20A Ul Bail Min Water Coverage (ft)

= B -W - 19A 22.002 E

Vessel Zero 757.646 M-17 20B U2 Bail Min Water Coverage (ift)

= B -W - 19B 21.794 10 11 Spent Fuel Storage Water Level 21.333 TS 3.7.8 21A U1 Fuel Rod Water Coverage (ft)

(B - W) - (D + Q + R + K + M) 22.747 12 1

13 Lengths Inches Reference 21B U2 Fuel Rod Water Coverage (ft)

(B - W) -(D

+ T + U + K + M) 22.539 14 F Bottom of Active Fuel (BoAF) (VZ) 216.310 197R616 22A Ul Dropped Assembly Coverage (ft)

= 20A -0 21.537 15 G FTop of Active Fuel (ToAF) (VZ) 366.310 197R616 22B U2 Dropped Assembly Coverage (ft)

= 20B -0 21.329 1W H Assembly Overall Length 176.140 107E1592 25 Vessel Zero

= E 757.646 17 t ssembly "Below Seating Surface" 2.165 107E1592 27 ToAF in RPV

= E + G 788.172 18 J Assembly "Above Seating Surface" 173.975 107E1592 28 Top of Bail Handle in RPV

=E + G + P 789.718 19 K Assembly Seating Surface to BoAF 5.261 107E1592 31 Full UP Bottom of Assembly

= A -X - H 820.905 L

Assembly Seating Surface to ToAF 155.261 107E1592 32A Ul SFP Drop Height (ft)

= 31 - (D + Y + R + J) 0.990 21 M Fuel Pin Length (exclusive of End Caps) 159.770 GNF 234C5304 32B U2 SFP Drop Height (ft)

= 31 - 19B 0.615 2T N Plenum Length 9.770 GNF 234C5304 33A Ul Assembly Lying on Bail Handles

= 19A + 0 820.546 23 0 Assembly Width 5.576 GNF 103E1385 33B U2 Assembly Lying on Bail Handles

= 19B + 0 820.754 24P Assembly ToAF to Top of Bail 18.550 107E1592 34A U1 Top of Fuel Rod

= 19A - P + N 819.350 2

Unit 1 Rack Maximum Pedestal Height 9.250 Holtec 903 34B U2 Top of Fuel Rod

= 19B - P + N 819.558 26 R Unit 1 Rack Bottom Plate 0.750 Holtec 902 27 s Unit 1 Rack Height (Plate to Top) 167.750 Holtec 902 "Tech Spec 3.7.8 Requires a minimum of 21ft 4 inches of water coverage. Should use the minimum of 28 T Unit 2 Rack Maximum Pedestal Height 12.125 UST&D 8601-35 these numbers or 21.333 ft.

U29 Unit 2 Rack Bottom Plate 0.375 UST&D 8601-8 30 V Unit 2 Rack Height (Plate to Top) 167.750 UST&D 8601-2 20 Bail Min Water Coverage (ft) 21.333 W

Refuel Floor to Bottom of Scupper 17.000 S-784 21 Fuel Rod Water Coverage (ft)

= 20 + P - N 22.065 32 x Grapple Head to Top of Grapple Rail 96.000 TRM TSR 3.9.c.4 22 Dropped Assembly Coverage (ft)

= 20 - 0 20.868 33 Y Unit 1 Rack Minimum Pedestal Height 7.250 Holtec 903 1

1 Cale. No. L-003067, Rev. 1, Attachment E, Page E3 of E5

The analysis associated with the GE14 10x1O fuel is based on the NEDE-24011-P-A-US analysis and the known result for a 34 ft drop, namely 172 broken rods. {This analysis benchmarks an approach before applying it to an FHA over the spent fuel racks.) The general expression for the number of broken rods is:

drop height. (bundle + mast weight).

0.5 clad weight rods in d

fraction shared bundle weight - fuel weight drop energy with I

impacted fuel fraction of energy shared between clad and fuel l 1 1

1 total DroKen roas = dropped +

bundle dropped bundle 175 ft -lIbm rod energy per rod failure number broken due to initial impact bundle length. (mast weight + 0.5.bundle weight).

0.5 fraction shared with impacted fuel clad weight bundle weight - fuel weight fraction of energy shared between clad and fuel

+

175 ft -Ibm rod energy per rod failure number broken due to secondary impact where:

Drop Height

=34 ft Bundle Length

=160 in Mast Weight [Wet]

=619 lbs Bundle Weight [Wet]

=568 lbs Cladding Weight

=100.9 lbs Total assembly Weight [Dry]

=6451bs Total Pellet Weight

=455 lbs Energy per rod failure

=175 ft-Ibm/rod Fraction of Energy Absorbed by Clad:

100.9 lbm 0.531 645 ibm - 455 ibm

[Ref. 11]

[Ref. 11]

[Ref. 11]

[Ref. 12]

[Ref. 12]

[Ref. 12]

[Ref. 10]

[Ref. 10,12]

Inserting the above values, one obtains the following:

160in. (619 + 0.5. 568). 0.5

  • 0.529 12in 34fti.(6191b+ 568lbm).0.5.0.529 f12 92rods +

+=-172rods dropped 175 ft-Ibm 175 ft-Ibm bundle rod rod initial impact-62 rods seconday impacted 8 rods Thus, in the case of a 1 Ox1 0 bundle the number of failed rods would be 92 from the impacting (dropped) assembly 62 (H/34) from the impacted assemblies when H is the height of the drop 18 from the second impact.

Bounding FHA Assessment Caic. No. L-003067, Rev. 1, Attachment E, Page E4 of E5

This analysis is for the GE14 1Ox10 fuel dropped over the spent fuel storage racks. The drop height is rounded up from 3.478 feet to 4.0 feet even. Again, the general expression for the number of broken rods is:

drop height. (bundle + mast weight).

0.5 clad weight rods in drop energy fraction shared bundle weight - fuel weight

dropped +

impacted fuel fraction of energy shared between clad and fuel bundle 175 ft -lbm total broken rods =

dropped bundle rod energy per rod failure number broken due to initial impact bundle length. (mast weight + 0.5 bundle weight).

0.5 clad weight fraction shared bundle weight - fuel weight with impacted fuel fraction of energy shared between clad and fuel

+

175ft -Ibm rod energy per rod failure number broken due to secondary impact where:

Drop Height Bundle Length Mast Weight [Wet]

Bundle Weight [Wet]

Cladding Weight Total assembly Weight [Dry]

Total Pellet Weight Energy per rod failure

=4 ft

=160 in

=619 lbs

=568 lbs

=100.9 lbs

=645lbs

=455 lbs

=175 ft-Ibm/rod Rounded up from pg E-3 value

[Ref. 11]

[Ref. 11]

[Ref. 12]

[Ref. 12]

[Ref. 12]

[Ref. 10]

[Ref. 10,12]

Fraction of Energy Absorbed by Clad:

100.9 Ibm

=0.531 645 Ibm - 455 Ibm Inserting the above values, one obtains the following:

12in ft 92rods + 4ft.(6191b + 568 ibm). 0.5-0.529 dropped 175 fibm bundle rod initial impacta7 rods

= 117 rods 175 ft - lbm rod seconday impactal 8 rods Thus, in the case of a 1 Ox1 0 bundle the number of failed rods would be 92 7

18 from the impacting (dropped) assembly from the impacted assemblies from the second impact.

Bounding FHA Assessment Calc. No. L-003067, Rev. 1 Attachment E, Page E5 of E5

Computer Disclosure Sheet Discipline Nuclear Client:

Exelon Corporation Date: October 2008 Project:

LaSalle County Station Job No.

28062-LAS0152 Program(s) used Rev No.

Rev Date Calculation Set No.: L-003067, Rev. 1 Attachment E spreadsheet N/A Status

[ ] Prelim.

[X] Final

[

] Void.

WGI Prequalification

[ ] Yes

[X] No Run No.

==

Description:==

Analysis

Description:

Spreadsheet used to perform water coverage and fuel damage assessment for FHA, 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:

Run by: H. Rothstein Checked by: P. Reichert F44 A

Approved by: H. Rothstein 1 Z~/.

Remarks: WGI Form for Computer Software Control This spreadsheet is relatively straight-forward and was hand checked. Attachment E includes the spreadsheet in both normal and formula display mode so it is completely documented.

L-003067, Rev. 1, Attachment F, Page F I of F2

Computer Disclosure Sheet Discipline Nuclear Client:

Exelon Corporation Date: October 2008 Project:

LaSalle County Station FHA AST Job No.

28062-LAS01 52 Program(s) used:

Rev No.

Rev Date Calculation Set No.: LS-003067, Rev. 1 RADTRAD 3.03 Runs in Att. 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 FHA 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:

Run by: W. Golden

/ 0/boo/

Checked by: P. Reichert Lb (1Y Approved by: H. Rothstein 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 LaSalle County Generating Station. Both were also hand checked for accuracy.

L-003067, Rev. 1, Attachment F, Page F2 of F2 A